TECHNICAL BASIS

CONTAINMENT INTEGRITY SIGNIFICANCE DETERMINATION PROCESS

(IMC 0609, App H)

FOR TYPE A AND TYPE B FINDINGS

FULL POWER AND SHUTDOWN OPERATIONS

TABLE OF CONTENTS

LIST OF TABLES AND FIGURES [3](#__RefHeading___Toc193082658)

Technical Basis for Type A and Type B Findings - Full Power Operations
[4](#__RefHeading___Toc193082659)

1.0 Type A Findings [4](#__RefHeading___Toc193082660)

1.1 Accident Sequence Screening Criteria for Type A Findings
[5](#__RefHeading___Toc193082661)

BWR Accidents Important to LERF [5](#__RefHeading___Toc193082662)

PWR Accidents Important to LERF [5](#__RefHeading___Toc193082663)

1.2 LERF Factors for LERF-Significant Sequences
[6](#__RefHeading___Toc193082664)

ISLOCA [6](#__RefHeading___Toc193082665)

ATWS [6](#__RefHeading___Toc193082666)

Steam Generator Tube Ruptures (SGTR) [7](#__RefHeading___Toc193082667)

Mark I Plants: Scenarios with the RCS at High Pressure and Drywell Floor
Flooded/Dry [7](#__RefHeading___Toc193082668)

Mark II Plants: Scenarios with the RCS at High Pressure
[7](#__RefHeading___Toc193082669)

Mark III Plants: Scenarios with RCS at High Pressure and SBO Sequences
with RCS at High or Low Pressure [7](#__RefHeading___Toc193082670)

Ice Condenser Plants: Station Blackout Scenarios
[8](#__RefHeading___Toc193082671)

2.0 Type B Findings [8](#__RefHeading___Toc193082672)

2.1 Phase 1 Screening Decision Criteria for Type B Findings
[10](#__RefHeading___Toc193082673)

2.2 LERF Factors for LERF-Significant SSCs
[10](#__RefHeading___Toc193082674)

Containment Penetration Seals, Isolation Valves, Vent and Purge Systems
[11](#__RefHeading___Toc193082675)

Systems Important to Suppression Pool Integrity - BWR Only
[13](#__RefHeading___Toc193082676)

Main Steam Isolation Valves - BWR Mark I and II Only
[14](#__RefHeading___Toc193082677)

Drywell/Containment Sprays - BWR only [14](#__RefHeading___Toc193082678)

Hydrogen Igniters - BWR Mark III Plants
[15](#__RefHeading___Toc193082679)

Hydrogen Igniters - PWR Ice Condenser Plants
[16](#__RefHeading___Toc193082680)

Ice Condenser Integrity - PWR Ice Condenser Plants Only
[17](#__RefHeading___Toc193082681)

Technical Basis for Type A and Type B Findings - Shutdown Operation
[18](#__RefHeading___Toc193082682)

Definitions [18](#__RefHeading___Toc193082683)

Plant Operational States [18](#__RefHeading___Toc193082684)

Time Windows [19](#__RefHeading___Toc193082685)

Containment Status [21](#__RefHeading___Toc193082686)

Type A Findings at Shutdown - Phase 1 Screening
[21](#__RefHeading___Toc193082687)

Type A Findings at Shutdown - Phase 2 Assessment
[21](#__RefHeading___Toc193082688)

Type B Findings at Shutdown - Phase 1 [22](#__RefHeading___Toc193082689)

Type B Findings at Shutdown - Phase 2 [23](#__RefHeading___Toc193082690)

LIST OF TABLES AND FIGURES[]{#__RefHeading___Toc193082658 .anchor}

List of Tables

[Table 1A.2 Phase 2 Risk Significance - Type B Findings at Full Power
11](#Table_1A2)

[Table 1A.3 Containment Leakage Rate Significant to LERF 13](#Table_1A3)

[Table B-3 POSs Baseline Annualized CDF vs. Plant Mitigation Capability
24](#Table_B3)

List of Figures

[Figure 1B.1 POS and Time Windows for BWRs at Shutdown 20](#Table_1b1)

[Figure 1B-2 POS and Time Windows for PWRs at Shutdown 20](#Table_1b2)

**Technical Basis for Type A and Type B Findings - Full Power
Operations**[]{#__RefHeading___Toc193082659 .anchor}

**Introduction**

The screening criteria and technical basis for the Phase 2 assessment
guidance for Type A and Type B findings are discussed below, for each of
the five reactor and containment design combinations. The screening
criteria and the numerical factors are based largely on NUREG-1765,
"Basis Document for Large Early Release Frequency (LERF) Significance
Determination Process," (Reference 6), to which the reader is referred
for more detail.

**1.0 Type A Findings**[]{#__RefHeading___Toc193082660 .anchor}

A subset of the findings processed through the core damage frequency
(CDF) based significance determination process (SDP) can also affect
LERF. Findings of this nature are classified as "Type A" in Figure 4.1.
Guidance for the assessment of Type A findings is presented in the form
of screening criteria to identify those accident sequences that
contribute to LERF. These screening criteria are based on the
characteristics of the sequence (see Table 5.1). If a finding is
evaluated to affect an accident sequence that contributes to LERF, that
sequence is examined further in a Phase 2 assessment.

The containment SDP (Appendix H to IMC 609) is based on the
identification of a factor that represents the conditional probability
of a large early release for each LERF-contributing sequence. The factor
is dependent on the nature of the accident sequence (see Table 5.2). The
change in LERF is derived from the changes in the accident sequence
frequencies in the following way:

![](media/image1.wmf) ![](media/image2.wmf) ![](media/image3.wmf) (1)

where the index i runs over the accident sequences that contribute to
LERF.

Only certain core damage sequences represent a challenge to containment
and are different for each reactor/containment type combination. The
sequences are summarized below. Core damage sequences that do not pose a
threat to containment integrity would not result in a significant
increase in LERF and are therefore not included in Equation (1) above.

In general, the information used to address containment performance for
Type A findings has been derived from the NUREG-1150 study (Reference 7)
and supplemented by studies addressing resolution of selected severe
accident issues. The factors for Type A findings given below are based
on average values for each reactor/containment type combination. Plant
specific values may be used to obtain the risk significance of a finding
for individual plants when performing a Phase 3 assessment.

**1.1 Accident Sequence Screening Criteria for Type A
Findings**[]{#__RefHeading___Toc193082661 .anchor}

The CDF-based SDP will result in a risk categorization for the finding,
and an assessment of the types of accident sequences whose frequency
would increase as a result of the finding. Initially all inspection
findings pass through the screening decision criteria shown in Table
5.1. The finding is assessed for each of the accident categories
contributing to the risk significance ranking from the CDF-based SDP for
various reactor and containment type combinations. The outcome of the
screening process shown in Table 5.1 is whether or not a Phase 2
assessment needs to be performed with respect to LERF.

**[BWR Accidents Important to
LERF]{.underline}**[]{#__RefHeading___Toc193082662 .anchor}

For BWR plants with Mark I and Mark II containments, findings related to
interfacing system loss of coolant accidents (ISLOCA), anticipated
transient without scram (ATWS), and accidents resulting in high reactor
coolant system (RCS) pressure (i.e., transients and small break LOCA)
need to be further evaluated in Phase 2. For Mark I plants, core damage
accidents that involve a dry drywell floor at vessel breach regardless
of whether the RCS is at low or high pressure also need to be evaluated
in Phase 2 as indicated in Note 2 to Table 5.1. For BWR Mark III plants,
findings related to ISLOCA, transients, small break LOCAs, and station
blackout (SBO) categories need to be further evaluated in Phase 2.

ATWS sequences are [not]{.underline} an important contributor to LERF
for BWRs with Mark III containment. Containment failure from ATWS
sequences occurs due to gradual over-pressurization of containment prior
to core damage. However, these sequences leave the drywell and
suppression pool intact, hence the releases are scrubbed by the pool and
a large early release does not occur.

**[PWR Accidents Important to
LERF]{.underline}**[]{#__RefHeading___Toc193082663 .anchor}

For PWR plants with large dry and sub-atmospheric containments, findings
related to the accident categories ISLOCA and steam generator tube
rupture (SGTR) need to be further evaluated in Phase 2. For the PWR
plants with ice condenser containments, findings related to ISLOCA,
SGTR, and SBO accident categories need evaluation in Phase 2.

In PWRs with ice condenser containments, severe accident studies
indicate that the most significant factor is the availability of
hydrogen igniters and the ice condenser to mitigate severe accidents. If
the igniters are available (i.e., non-SBO accidents), the conditional
early containment failure probability is less than 0.1 even during
accidents that leave the RCS at high pressure. However, if the igniters
are not available (as they would be in an SBO), NUREG/CR6527 indicates a
conditional containment failure probability (CCFP) close to 1.0.

ATWS sequences are [not]{.underline} significant contributors to LERF
for PWRs. During ATWS in PWRs, the containment pressure increases slowly
so it can be considered a late failure mode. Therefore, the risk
significance determined by the CDF based SDP alone for ATWS events in
PWRs is appropriate.

High and low pressure core damage sequences (in which the containment is
not bypassed) are [not]{.underline} significant contributors to LERF for
PWRs with large dry and sub-atmospheric containments. An important
insight from the IPE program and other PRAs is that the conditional
probability of early containment failure is less than 0.1 on the
occurrence of core damage accident scenarios that leave the RCS at high
pressure. If the RCS is depressurized the probability of early
containment failure is less than 0.01.

**1.2 LERF Factors for LERF-Significant
Sequences**[]{#__RefHeading___Toc193082664 .anchor}

LERF factors for sequence types affecting LERF associated with Type A
findings are shown in Table 5.2 and discussed individually below.

**[ISLOCA]{.underline}**[]{#__RefHeading___Toc193082665 .anchor}

The ISLOCA scenario occurs when isolation valves between the high
pressure RCS and a low pressure system fail causing a LOCA outside
containment. If core damage occurs, the release path can bypass
containment and cause a large release to the environment. For BWRs and
PWRs, an ISLOCA is potentially a high consequence accident sequence
since the containment is bypassed. Although some fission product holdup
and scrubbing would occur along the release path, this depends on the
break location and plant-specific features and is not credited in this
SDP. Thus, the factor is equal to 1.0 for this accident class.

**[ATWS]{.underline}**[]{#__RefHeading___Toc193082666 .anchor}

ATWS accident sequences are significant contributors to LERF for BWRs
with Mark I and Mark II containment designs. These accident sequences
result in an energy input to containment that cannot be adequately
removed by the normal containment heat removal systems. This results in
a rapid pressure rise that may cause the containment to fail before or
shortly after core damage. If the suppression pool is bypassed, a large
release may occur. The NRC review of individual plant examinations
(IPEs), NUREG-1560, Volume 2, indicated that the significance of ATWS
events in the various IPEs depends on plant-specific features such as
the ability of pumps to operate with saturated water, and on assumptions
about the power level, the period in the fuel cycle when the event
occurs, and the effectiveness of operator response.

For Mark I and II containments, the Factor which represents conditional
probability of early containment failure and suppression pool bypass due
to ATWS sequences is estimated by the equation below:

![](media/image4.wmf)

where: Factor is the multiplier on the CDF

> C~PEF~ is the conditional probability of early containment failure
> given ATWS, and
>
> C~PPB~ is the conditional probability of pool bypass given early
> containment failure.

Based on data from the IPE program, averaged across Mark I plants,
C~PEF~ = 0.6 and C~PPB~ = 0.5. This results in a Factor of 0.3 for ATWS
sequences. The corresponding average values for Mark II containments are
C~PEF~ = 0.6 and C~PPB~ = 0.7 resulting in a Factor of 0.4 .

**[Steam Generator Tube Ruptures
(SGTR)]{.underline}**[]{#__RefHeading___Toc193082667 .anchor}

SGTRs are initiating events that can lead, after further failures, to
core damage. If core melt occurs, and the secondary side is open, the
release path can bypass containment. It was conservatively assumed that
all SGTR core damage sequences would release sufficiently large amounts
of radioactive material to the atmosphere outside containment that, for
SDP purposes, would be treated as sequences important to LERF. As the
containment is bypassed for SGTR sequences, the conversion factor for
Type A findings is 1.0 for this accident class.

**[Mark I Plants: Scenarios with the RCS at High Pressure and Drywell
Floor Flooded/Dry]{.underline}**[]{#__RefHeading___Toc193082668 .anchor}

Information provided in NUREG/CR-6595 (Reference 9) for Mark I plants
indicates that if the RCS is at high pressure at vessel breach and the
drywell floor is dry, then the conditional probability of containment
failure is 1.0. If the RCS is at high pressure at vessel breach and the
drywell floor is flooded, then the average conditional probability of
containment failure in Mark I plants is 0.6. NUREG/CR-5423 (Reference
10) documented an analysis to evaluate the probability of containment
liner melt through in Mark I plants for sequences where the vessel fails
at low pressure. This analysis considered both oxidic and metallic pours
with and without water on the drywell floor. If there is no water on the
drywell floor, the conditional containment failure probability of the
containment from corium attack was estimated to be 1.0 for oxidic melts
and 0.63 for metallic melts. Given the uncertainty in core melt
progression and molten debris composition, for sequences involving
drywell floor at vessel breach, a factor of 1.0 is conservatively
assumed in the SDP, regardless of RCS pressure. The same study indicates
that if the reactor coolant system pressure is low and the drywell floor
is flooded then the conditional containment failure probability and
factor is less than 0.1.

**[Mark II Plants: Scenarios with the RCS at High
Pressure]{.underline}**[]{#__RefHeading___Toc193082669 .anchor}

Based on data provided in NUREG/CR-6595 (Reference 9) for Mark II
plants, the likelihood of containment failure given a vessel breach is
approximately 0.3 if the RCS is at high pressure. This probability is
relatively independent of whether the pedestal or drywell floor is
flooded or not for the high pressure scenarios.

**[Mark III Plants: Scenarios with RCS at High Pressure and SBO
Sequences with RCS at High or Low
Pressure]{.underline}**[]{#__RefHeading___Toc193082670 .anchor}

Mark III plants have a drywell and suppression pool that is totally
enclosed within a containment structure. Thus, both the containment and
the drywell have to fail or be bypassed for a large release to occur.
The NUREG-1150 (Reference 7) study indicates that the likelihood that an
accident sequence contribute to LERF depends on two factors: (1) whether
the accident is a SBO sequence (hydrogen igniters not operating and
containment sprays unavailable) and (2) the pressure in the RCS at the
time of vessel breach. In a SBO sequence, the conditional probability of
a large early release is approximately 0.2 (due to hydrogen combustion)
regardless of whether the RCS is at high or low pressure. If the RCS is
at high pressure, the conditional probability of failure of both
containment and the drywell is also about 0.2 (due to overpressure
failure). Thus, for findings related to any SBO or transient with high
RCS pressure the Factor would be 0.2.

**[Ice Condenser Plants: Station Blackout
Scenarios]{.underline}**[]{#__RefHeading___Toc193082671 .anchor}

Like BWR Mark III plants, ice condenser containments rely on AC-powered
glow plug igniters to control hydrogen released during core damage
accidents. A recent detailed study of severe accident phenomena in ice
condenser plants NUREG/CR-6427 (Reference 11) indicates that the most
significant challenge to containment integrity arises from SBO sequences
in which the igniters are not available. The average CCFP for SBO
sequences from hydrogen combustion across five ice condenser plants is
about 0.6 (range from 0.22 to 0.97). For a screening SDP, this Factor is
conservatively assessed a value of 1.0.

**2.0 Type B Findings**[]{#__RefHeading___Toc193082672 .anchor}

Findings that have no impact on the determination of the ΔCDF but are
potentially important to ΔLERF determinations are classified as Type B
findings. Type B findings are fundamentally different from Type A
findings in that they are not processed through the CDF based SDP and
have to be allocated significance categories based only on LERF
considerations. Table 1A-1 provides a list of SSCs associated with
maintaining containment integrity in different containment types that
were reviewed to determine if ΔLERF would be affected if the SSCs were
found to be degraded.

As the containment function may be compromised for a Type B finding it
can potentially affect either all core damage accidents or a subset of
those accidents. Baseline CDFs were assumed in order to simplify the
calculation of the change in risk for this type of finding. The baseline
CDFs assumed were 10^-4^/reactor year for PWRs and 10^-5^/ reactor-year
for BWRs. Assuming a baseline CDF is a limitation that has to be
recognized in the light of the relatively wide ranges associated with
Plant-Specific CDF estimates.

The risk significance categories are obtained using the following
relationship:

ΔLERF = F~B~ x F~R~ x CDF x (multiplier for the duration of degraded
condition)

where:

> F~B~ = increase in the conditional large early release probability
> resulting from the finding, and
>
> F~R~ = fraction of the CDF for which the containment failure mode
> affected by the finding impacts LERF.

In the above relationship the duration of the degraded condition is a
simple multiplier for the following three time periods:

[Duration Multiplier]{.underline}

\>30 days 1.0

30-3 days 0.1

\< 3 days 0.01

For Type B findings, the factor F~B~ is the difference between the
conditional probability of large early release assuming complete failure
of the SSC and the original conditional probability of large early
release assumed in the baseline risk estimate. F~B~ is multiplied by the
fraction F~R~ of CDF that pertains to the phenomena that impact (or are
impacted by) the failure of the SSC under consideration.

For example, suppose a finding in a BWR with a Mark III containment or a
PWR with an ice condenser containment indicates that a significant
number of hydrogen igniters are not operable. Potentially, such a
finding implies that the containment is vulnerable to failure from a
hydrogen deflagration or detonation after a core damage accident.
However, the igniters require AC power to operate and would therefore
not be available anyway in a SBO accident. Hence, the risk significance
of the finding pertains only to the non-SBO portion of the total CDF.
Generic values for these parameters are provided in the phase 2 guidance
contained herein. In a phase 3 evaluation, the plant-specific CDF for
the major accident classes, as well as a more precise estimate of the
multiplier (if the actual duration can be established), should be taken
into consideration in making a final significance determination.

If F~B~ is 1.0 and duration of the degraded condition is \>30 days, the
implication is that ΔLERF is equivalent to either the total CDF or the
fraction F~R~ of CDF that is relevant to the containment function that
is compromised. If the total CDF is involved, then, under these
circumstances (refer to Table 1.1), the risk significance based on LERF
is red for both PWRs and BWRs. As the product of F~B~ and F~R~
decreases, the influence of the containment SSC on the determination of
LERF also decreases correspondingly. When this product reaches 0.1, the
implication is that the significance category becomes one order of
magnitude lower, i.e. the LERF significance is equivalent to the CDF
significance.

The bases for the significance categories for each of the five
reactor/containment combinations are provided in summary below. The data
used to establish the risk significance of Type B findings is based
mainly on the results of the IPE program reported in NUREG-1560
(Reference 8). Containment related findings that are not addressed in
this LERF SDP are not expected to be risk significant with respect to
LERF.

The characteristics of the severe accidents that contribute to LERF and
those SSCs important to maintain containment integrity are plant
specific and depend upon the containment design (i.e., large containment
volume versus the various pressure suppression designs). The screening
criteria are therefore presented below in the context of five
containment designs: BWRs with Mark I, II, or III containments, PWRs
with large dry and sub-atmospheric containments, and PWRs with ice
condenser containments.

**2.1 Phase 1 Screening Decision Criteria for Type B
Findings**[]{#__RefHeading___Toc193082673 .anchor}

Table 1A-1 lists the containment SSCs that were reviewed to determine
the impact on ΔLERF if they were found to be in a degraded. Table 6.1
provides the Phase 1 screening criteria for Type B findings for all
reactor and containment type combinations. Findings involving
containment SSCs that do not appear in Table 6.1 may be considered to be
screened out in Phase 1. Main steam isolation valve (MSIV) leakage as a
containment SSC failure mode is screened out for BWR Mark III plants
since these containments have a safety-grade main steam shut-off valve
(MSSV) that is a relatively slow closing low leakage valve. Thus,
leakage past the MSIV in a Mark III plant would likely be stopped by the
MSSV.

**2.2 LERF Factors for LERF-Significant
SSCs**[]{#__RefHeading___Toc193082674 .anchor}

Each Type B finding that was not screened out in Phase 1 requires a
Phase 2 assessment. The risk significance of the various findings, i.e.,
color assignment, are summarized in Table 6.2 in terms of the duration
that the condition was estimated to exist. Table 1A.2 provides details
of the values of F~B~ and F~R~ for different reactor and containment
type combinations that were used to generate the risk significance color
assignments. The bases for assigning these values are discussed below.

The values of F~B~ and F~R~ given in the tables below are average values
for each reactor/containment type combination. Plant specific values of
these parameters may be substituted to obtain the risk significance of a
finding for individual plants when performing a Phase 3 assessment.

+-------+-------------------------------+-----+---+-----+-----+-----+
| []{   |                               |     |   |     |     |     |
| #Tabl |                               |     |   |     |     |     |
| e_1A2 |                               |     |   |     |     |     |
| .anch |                               |     |   |     |     |     |
| or}** |                               |     |   |     |     |     |
| Table |                               |     |   |     |     |     |
| 1A.2  |                               |     |   |     |     |     |
| Phase |                               |     |   |     |     |     |
| 2     |                               |     |   |     |     |     |
| Risk  |                               |     |   |     |     |     |
| Si    |                               |     |   |     |     |     |
| gnifi |                               |     |   |     |     |     |
| cance |                               |     |   |     |     |     |
| -Type |                               |     |   |     |     |     |
| B     |                               |     |   |     |     |     |
| Findi |                               |     |   |     |     |     |
| ngs** |                               |     |   |     |     |     |
| **at  |                               |     |   |     |     |     |
| Full  |                               |     |   |     |     |     |
| Po    |                               |     |   |     |     |     |
| wer** |                               |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| Re    | Finding                       | F   | F | R   |     |     |
| actor |                               | ~B~ | ~ | isk |     |     |
| and   |                               |     | R | Sig |     |     |
| C     |                               |     | ~ | nif |     |     |
| ontai |                               |     |   | ica |     |     |
| nment |                               |     |   | nce |     |     |
| Type  |                               |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       |                               |     |   | \>  | 3   | \<  |
|       |                               |     |   | 30  | 0 - | 3   |
|       |                               |     |   | d   | 3   | d   |
|       |                               |     |   | ays | d   | ays |
|       |                               |     |   |     | ays |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| BWR   | Leakage from drywell to       | 0.9 | 1 | Yel | Wh  | Gr  |
| Mark  | environment \> 100%           |     | . | low | ite | een |
| I     | volume/day through            |     | 0 |     |     |     |
|       | containment penetration       |     |   |     |     |     |
|       | seals, isolation valves or    |     |   |     |     |     |
|       | vent and purge systems        |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Failure of systems/components | 0.2 | 1 | Yel | Wh  | Gr  |
|       | critical to suppression pool  |     | . | low | ite | een |
|       | integrity or scrubbing        |     | 0 |     |     |     |
|       | (vacuum breakers or other     |     |   |     |     |     |
|       | bypass mechanisms)            |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Main steam isolation valve    | 0.2 | 0 | Yel | Wh  | Gr  |
|       | leakage \>10,000 scfh through |     | . | low | ite | een |
|       | the best sealing valve in any |     | 5 |     |     |     |
|       | steam line                    |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Drywell sprays unavailable    | 0.7 | 1 | Yel | Wh  | Gr  |
|       |                               |     | . | low | ite | een |
|       |                               |     | 0 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| BWR   | Leakage from drywell to       | 0   | 1 | Yel | Wh  | Gr  |
| Mark  | environment \> 100%           | .95 | . | low | ite | een |
| II    | volume/day through            |     | 0 |     |     |     |
|       | containment penetration       |     |   |     |     |     |
|       | seals, isolation valves or    |     |   |     |     |     |
|       | vent and purge systems        |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Failure of systems/components | 0.1 | 1 | Yel | Wh  | Gr  |
|       | critical to suppression pool  |     | . | low | ite | een |
|       | integrity or scrubbing        |     | 0 |     |     |     |
|       | (vacuum breakers or other     |     |   |     |     |     |
|       | bypass mechanisms)            |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Main steam isolation valve    | 0.7 | 0 | Yel | Wh  | Gr  |
|       | leakage \>10,000 scfh through |     | . | low | ite | een |
|       | the best sealing valve in any |     | 7 |     |     |     |
|       | steam line                    |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Drywell sprays unavailable    | 0.2 | 0 | Wh  | Gr  | Gr  |
|       |                               |     | . | ite | een | een |
|       |                               |     | 3 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| BWR   | Leakage from wetwell to       | 0.  | 1 | Yel | Wh  | Gr  |
| Mark  | environment \>1000%           | 975 | . | low | ite | een |
| III   | containment volume/day        |     | 0 |     |     |     |
|       | through containment           |     |   |     |     |     |
|       | penetration seals, isolation  |     |   |     |     |     |
|       | valves or vent and purge      |     |   |     |     |     |
|       | systems                       |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Failure of systems/components | 0   | 0 | Yel | Gr  | Yel |
|       | critical to suppression pool  | .23 | . | low | een | low |
|       | integrity or scrubbing        |     | 1 |     |     |     |
|       | (vacuum breakers or other     |     |   |     |     |     |
|       | bypass mechanisms)            |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Failure of multiple igniters  | 0   | 0 | Wh  | Gr  | Yel |
|       | such that coverage is lost in | .19 | . | ite | een | low |
|       | two adjacent compartments     |     | 2 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Containment sprays            | 0.  | 1 | Wh  | Gr  | Gr  |
|       | unavailable                   | 025 | . | ite | een | een |
|       |                               |     | 0 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| PWR   | Leakage from containment to   | 0.9 | 1 | Red | Yel | Wh  |
|       | environment \> 100%           |     | . |     | low | ite |
| Large | containment volume/day        |     | 0 |     |     |     |
| Dry   | through containment           |     |   |     |     |     |
| and   | penetration seals, isolation  |     |   |     |     |     |
| Sub-A | valves or vent and purge      |     |   |     |     |     |
| tmosp | systems                       |     |   |     |     |     |
| heric |                               |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
| PWR   | Leakage from containment to   | 0.9 | 1 | Red | Yel | Wh  |
|       | environment \> 100%           |     | . |     | low | ite |
| Ice   | containment volume/day        |     | 0 |     |     |     |
| Cond  | through containment           |     |   |     |     |     |
| enser | penetration seals, isolation  |     |   |     |     |     |
|       | valves or vent and purge      |     |   |     |     |     |
|       | systems                       |     |   |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Blockage of more than 15% of  | 0.9 | 1 | Red | Yel | Wh  |
|       | the flow passage into or      |     | . |     | low | ite |
|       | through the ice bed           |     | 0 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+
|       | Failure of multiple igniters  | 0.9 | 0 | Red | Yel | Wh  |
|       | such that coverage is lost in |     | . |     | low | ite |
|       | two adjacent compartments     |     | 9 |     |     |     |
+-------+-------------------------------+-----+---+-----+-----+-----+

**[Containment Penetration Seals, Isolation Valves, Vent and Purge
Systems]{.underline}**[]{#__RefHeading___Toc193082675 .anchor}

The risk significance of a finding relative to a loss of containment
penetration seals, isolation valves or vent and purge systems is
assessed in terms of the leakage rate from containment (drywell to
environment for BWR Mark I and II, wetwell to environment for BWR Mark
III, and containment airspace to environment for PWR plants). Several
studies, including NUREG/CR-4330, NUREG-1493, and NUREG/CR-6418, have
been performed to determine the risk significance of various levels of
containment leakage. While the results vary by plant and containment
type, a containment leak rate of about 100% containment volume/day
appears to constitute an approximate threshold beyond which the release
may become significant to LERF. In Mark III plants, however, the impact
of suppression pool decontamination factors (DFs) has to be taken into
account when considering the leakage from these containments.
Conservatively, a DF of 10 has historically been used to represent a
pool over the entire accident period. In terms of an "early release"
this factor is extremely conservative. Including this DF to determine
the containment leakage criterion of importance to LERF would imply a
wetwell to environment leak rate of about 1000% containment volume/day.

Design basis leakage from containment is determined by regulatory
requirements to assure that the containment leakage will be below the
design basis (or maximum allowable) leak rate L~a~ set by the 10CFR Part
100 dose limits that are incorporated in the plant technical
specifications. While individual plants have been licensed with various
different values of L~a~, typical values of L~a~ are 0.1% containment
volume/day for PWRs and 0.5% containment volume/day for BWRs. Thus a
LERF significant leakage rate from containment of 100% containment
volume/day would correspond to about 1000 L~a~ for PWRs and 200 L~a~ for
BWR Mark I and Mark II plants. In Mark III plants, the LERF significant
leakage rate of 1000% containment volume/day would correspond to 2000
L~a~.

The comparison between % containment volume/day and L~a~ is shown in
Table 1A.3 for various reactor/containment type combinations. The 100
volume percent per day leakage rate is approximately equivalent to a
hole size in containment of 2.5-3.0 inches in diameter for PWRs with
large dry containments, 2 inches for PWRs with ice condenser
containments, 1 inch for BWRs with Mark I or II containments, and 2.5
inches for BWRs with Mark III containments.

+----------------------+----------------------+-----------------------+
| []{#Table_1A3        |                      |                       |
| .anchor}**Table 1A.3 |                      |                       |
| Containment Leakage  |                      |                       |
| Rate Significant to  |                      |                       |
| LERF**               |                      |                       |
+----------------------+----------------------+-----------------------+
| **R                  | **LERF-Significant   |                       |
| eactor/Containment** | Containment Leakage  |                       |
|                      | Rate**               |                       |
| **Type**             |                      |                       |
+----------------------+----------------------+-----------------------+
|                      | **% Containment      | **Design Basis        |
|                      | Volume/Day**         | Leakage Rate, L~a~**  |
+----------------------+----------------------+-----------------------+
| BWR Mark I and II    | 100                  | 200 L~a~              |
+----------------------+----------------------+-----------------------+
| BWR Mark III         | 1000                 | 2000 L~a~             |
+----------------------+----------------------+-----------------------+
| PWR - All            | 100                  | 1000 L~a~             |
+----------------------+----------------------+-----------------------+

Data generated in the IPE program and in published PRAs suggests that,
on average, about 0.1 of the CDF in BWR Mark I plants and 0.05 of the
CDF in Mark II plants consists of LERF. Thus, if a finding were to imply
the existence of a breach in the drywell pressure boundary that would
result in a LERF significant drywell leakage rate shown in Table 1A.3
above, the large release probability would increase to essentially 1.0.
The factor F~B~ is thus (1.0 - 0.1 = 0.9) for Mark I plants and (1.0 -
0.05 = 0.95) for Mark II plants. In Mark III plants, IPE information
indicates that, on average, about 25% of CDF consists of early
containment failure sequences and about 10% of early containment failure
sequences consist of large releases. Hence, on average, about 0.025 of
the CDF consists of LERF. The relatively low baseline value of LERF for
Mark III plants reflects the fact that containment failure has to be
coupled with loss of drywell integrity (i.e., a factor of 0.1) for a
large release to occur. If a finding reveals that the containment
leakage would exceed the LERF-significant leakage rate (1000% per day
for a Mark III, which includes a factor of 10 credit for suppression
pool scrubbing ), the conditional probability of a large early release
(given core damage) would approach 1.0 for all core damage events. The
factor F~B~ for a LERF significant containment leak in Mark III plants
is thus (1.0 - 0.025 = 0.975). As the above results are based on the
total assumed CDF for BWRs, the value of F~R~ is 1.0 for findings of
this type since essentially all core damage events would lead to a large
early release given a large containment leak.

For PWR plants, the conditional probability of early containment failure
and bypass given core damage is approximately 0.1. A finding that
implies the existence of a breach in containment that would result in a
LERF significant containment leak rate shown in Table 1A.3 above would
open a direct path to the environment, and hence the factor F~B~ is
(1.0 - 0.1 = 0.9) for PWR plants. As the above are based on the total
assumed CDF for PWRs the value of F~R~ is 1.0 for findings of this type.

**[Systems Important to Suppression Pool Integrity - BWR
Only]{.underline}**[]{#__RefHeading___Toc193082676 .anchor}

These findings pertain to failures of those systems that are important
for maintaining suppression pool integrity and preventing suppression
pool bypass in BWR plants, vacuum breakers in the wetwell airspace or
other bypass mechanisms. The values for F~B~ and F~R~ in Table 1A.2 were
obtained by reference to previous PRAs and in particular the results of
the IPE program, as documented in NUREG-1560 (Reference 8).

These values are described in the basis document for LERF SDP
(NUREG-1765). The approach adopted for determining the risk results was
to determine the average conditional probability of early containment
failure (given core damage) if the suppression pool was bypassed. The
following conditional early containment failure probabilities are
provided in NUREG-1765: 0.3 for Mark I containments, 0.15 for Mark II
containments and 0.25 for Mark III containments. The baseline average
conditional probability of large early release given core damage was
then determined for each containment, type. The baseline conditional
early containment failure probabilities reported in NUREG-1765 are: 0.1
for Mark I containments, 0.05 for Mark II containments and 0.025 for
Mark III containments. The finding is assumed to result in complete
suppression pool bypass and therefore the difference between the above
conditional probabilities (e.g., 0.3 - 0.1 = 0.2 for BWR Mark I plants,
0.15 - 0.05 = 0.1 for Mark II plants, and 0.25 - 0.025 = 0.23 for Mark
III plants) for each containment type represents F~B~ (i.e., the
increase in conditional probability of the finding). As the above
results are based on the total assumed CDF for BWRs, the value of F~R~
is 1.0 for findings of this type.

**[Main Steam Isolation Valves - BWR Mark I and II
Only]{.underline}**[]{#__RefHeading___Toc193082677 .anchor}

Excessive leakage from the MSIVs during a core melt accident in BWR Mark
I plants can lead to a release path that bypasses containment and can
impact LERF. Core melt accidents involving excessive MSIV leakage in
Mark I and Mark II plants with the RCS at high pressure are similar to
induced SGTRs in PWR plants. A finding in a BWR Mark I plant that
indicates excessive leakage from the MSIV in excess of 10,000 scfh
(standard cubic feet per hour) or an inability to quantify the leakage
rate would lead to the risk significance category indicated in Table
1A.2.

The values for F~B~ and F~R~ in Table 1A.2 were again obtained by
reference to previous PRAs and in particular the results of the IPE
program (as documented in NUREG-1560). In order to determine F~R~ it was
assumed that MSIV leakage is only risk significant for those accident
sequences in which the RCS is at high pressure during core melt. The IPE
results for BWR Mark I and Mark II plants were reviewed to determine the
fraction of the average CDF that consists of high RCS pressure sequences
for the two containment types of interest. These fractions were then
equated to F~R~. Based on the IPE results, F~R~ = 0.5 for Mark I plants
and F~R~ = 0.7 for Mark II plants.

The values for F~B~ were determined by obtaining the average conditional
probability of early failure and large release for high RCS pressure
sequences for each of the two containment types (i.e., 1.0 for Mark I
containments if the drywell floor is dry, 0.6 for Mark I containments if
the drywell floor is flooded and 0.3 for Mark II containments
independent of whether the drywell floor is dry or flooded). The finding
is assumed to result in a bypass probability of 1.0 for all high
pressure sequences and therefore the difference between the above
conditional probabilities for each containment type represents F~B~
(i.e., the increase in conditional probability of the finding). For Mark
II containments, F~B~ = 1.0 - 0.3 = 0.7. However, for Mark I
containments, F~B~ depends on whether or not the drywell floor is
flooded. If it is assumed that the drywell floor is flooded 50% of the
time, based on IPE results, the change in conditional probability can be
determined to be as follows:

F~B~ = 0.5 x (1.0-1.0) + 0.5 x (1.0-0.6) = 0.2.

**[Drywell/Containment Sprays - BWR
only]{.underline}**[]{#__RefHeading___Toc193082678 .anchor}

Using drywell sprays to flood the drywell floor during core melt
accidents in Mark I plants is an important strategy for reducing the
likelihood of LERF. In Mark II plants, containment sprays can impact
accidents that bypass the suppression pool by scrubbing the containment
atmosphere thus converting a potentially large release into a scrubbed
release. In Mark III plants, containment sprays can be used to provide
some fission product scrubbing.

As noted above, using the drywell sprays to flood the drywell floor
during core meltdown accidents in Mark I containments is an important
strategy for preventing liner melt-through and hence lowering the
likelihood of LERF. In addition to the status of the drywell floor
(flooded or dry), the pressure in the reactor vessel also has a
significant impact on the likelihood of LERF. Results generated in the
IPE program and reported in published PRAs suggests that for BWRs the
ratio of high to low pressure sequences is about 50/50. If a finding
were to imply that the drywell floor could not be flooded (using any
available means such as residual heat removal (RHR) pumps or
diesel-driven pumps) then the large release probability of 0.6 would
approach 1.0 for high pressure scenarios and would lie in the range of
\<0.1 to 1.0 for low pressure scenarios. F~B~ is therefore approximately
equal to {0.5 x (1.0 - 0.6) + 0.5 x (1.0 - \<0.1)} = 0.7 for findings of
this type. This assumption assumes that all liner melt-through failures
result in LERF and also neglects the effect of pool scrubbing for those
sequences in which the in-vessel release passes through the suppression
pool. As the above results are based on the total assumed CDF for BWRs
the value of F~R~ is 1.0 for findings of this type.

For those sequences where sprays are available, spray operation can be
beneficial, as mentioned above, for those accidents that challenge Mark
II containments. In these cases, timing is an important issue and there
is also a great deal of uncertainty in assessing the significance of
spray operability to reducing LERF. An approximate estimate of the
contribution of sprays can be made based on the results reported in
NUREG/CR-6595. The CCFP for low pressure sequences that fail containment
early is 0.3 without water on the drywell floor and 0.1 or less with
water on the drywell floor. (The CCFP for high pressure sequences that
fail containment early is unaffected by the presence of water). Assuming
conservatively, that water on the drywell floor can be ascribed to
successful spray operation, then based on the results reported in
NUREG/CR-6595, F~B~ = 0.3 - 0.1 = 0.2 for accident sequences in Mark II
plants in which the RCS is at low pressure. If we further assume, again
based on the IPE results, that about 30% of the CDF sequences are low
pressure sequences, then F~R~ = 0.3 in Mark II plants.

IPE information indicates that, on average, about 0.025 of the CDF in
Mark III BWRs contributes to LERF. Spray operation, if available and
effective, could potentially mitigate the fraction of the CDF that
contributes to LERF. The risk significance of a finding related to spray
operation therefore approaches conservatively a conditional probability
of 0.025 for a large early release assuming that the sprays are
effective for mitigating the LERF contributors. Therefore F~B~ = 0.025
in Mark III plants for a finding related to spray operation. As these
results are based on the assumed CDF for BWRs the value of F~R~ is 1.0
for findings of this type.

**[Hydrogen Igniters - BWR Mark III
Plants]{.underline}**[]{#__RefHeading___Toc193082679 .anchor}

These findings pertain to the operability of the glow plug hydrogen
igniter system (i.e., are all the glow plugs functioning) in BWR Mark
III plants under conditions where AC power is available. The loss off
all igniter coverage in two or more adjacent compartments is considered
potentially risk significant for hydrogen combustion events.

If a finding implies that a portion of the glow plug igniter system is
inoperable, the only impact on the probability of early containment
failure will be for non-SBO sequences in which the RCS is depressurized.
All SBO sequences (high pressure and low pressure) and non-SBO sequences
with the RCS at high pressure have a conditional probability of early
containment failure and simultaneous failure of the drywell close to
0.2. The probability of early containment failure from hydrogen
combustion events for non-SBO sequences with the RCS depressurized is
close to 0.01. This is because the igniters are operating and burn the
hydrogen at low concentrations as it released from the top of the
suppression pool with no containment challenging pressure spike. If a
portion of the igniters were not operating, the local concentration of
hydrogen would increase until it was ignited, either by a working
igniter elsewhere, or random ignition, e.g., static discharge.

In the extreme, if none of the igniters were operating, the probability
of early containment failure from non-SBO depressurized sequences would
approach 0.2 from hydrogen detonation or energetic deflagration.
Therefore F~B~ = 0.2 - 0.01 = 0.19 for findings that imply loss of
igniter system effectiveness.

F~R~ can be determined by assuming a contribution to the CDF for those
accidents where the igniters are normally effective in preventing
containment and drywell failure, namely non-SBO sequences with the RCS
at low pressure at the time of vessel breach (i.e., LOCAs and transients
with the RCS depressurized). Based on IPE results, the fraction of core
damage involving SBO is about 0.5, and the fraction of core damage
events involving low RCS pressure is about 0.4. Assuming that this
latter fraction applies to both SBO and non-SBO sequences, the
contribution to CDF from non-SBO sequences with RCS at low pressure is
about 0.5 x 0.4 = 0.2. Therefore, F~R~ = 0.2 for findings of this type
in Mark III plants.

**[Hydrogen Igniters - PWR Ice Condenser
Plants]{.underline}**[]{#__RefHeading___Toc193082680 .anchor}

These findings pertain to the operability of the glow plug hydrogen
igniter system (i.e., are all the glow plugs functioning) in PWR ice
condenser plants under conditions where AC power is available. The loss
off all igniter coverage in two or more adjacent compartments is
considered potentially risk significant for hydrogen combustion events.

If a finding implies that a portion of the glow plug igniter system is
inoperable, the only impact on the probability of early containment
failure will be for non-SBO sequences. (All SBO sequences have a
conditional probability of early containment failure close to unity).
The probability of early containment failure from non-SBO sequences from
hydrogen combustion events is zero because the igniters are operating
and burning the hydrogen at low concentrations as it enters the upper
compartment of the containment. If a portion of the igniters were not
operating, the local concentration of hydrogen would increase until it
was ignited, either by a working igniter elsewhere or by random
ignition(e.g., static discharge).

In the extreme, if none of the igniters were operating, the probability
of early containment failure from non-SBO sequences would approach 1.0
from hydrogen detonation or energetic deflagration. Since the average
conditional probability of early failure is about 0.1 in ice condenser
containments, F~B~ = 1.0 - 0.1 = 0.9.

However, the unavailability of igniters is a risk contributor only in
non-SBO accident sequences, hence we should consider only the non-SBO
fraction of core damage frequency. Based on the IPE database, the SBO
frequency as a fraction of CDF at the ice condenser plants ranges from
one percent to 21 percent with an average of approximately 10 percent.
Therefore F~R~ = 0.9 for findings of this type in PWR ice condenser
plants.

**[Ice Condenser Integrity - PWR Ice Condenser Plants
Only]{.underline}**[]{#__RefHeading___Toc193082681 .anchor}

Findings related to the integrity of the ice condenser involve failure
of some of the ice chest doors to open or a gross build-up of ice or
frost that results in a substantial blockage (15 percent or more of the
area ) of the flow path between ice baskets. There is also a possibility
of a significant loss of ice during or prior to an accident. If the
integrity of the ice condenser is lost, then it is assumed that
containment integrity is also lost. Since the average conditional
probability of early failure and bypass is about 0.1 in ice condensers
based on IPE data, F~B~ = 1.0 - 0.1 = 0.9 for loss of ice condenser
integrity. As these results are based on the assumed total CDF for PWRs,
the value of F~R~ is 1.0 for findings of this type.

**Technical Basis for Type A and Type B Findings - Shutdown
Operation**[]{#__RefHeading___Toc193082682 .anchor}

**Definitions**[]{#__RefHeading___Toc193082683 .anchor}

Similar to information used for Appendix G to IMC 609, the guidance for
assessing containment findings for plant shutdown divides an outage into
Time Windows (TWs) and Plant Operating States (POSs). Shutdown LERF
deficiencies are analyzed according to what TW and POS the finding
occurred in. For each TW and POS, the risk of that plant configuration
is assumed to stay constant.

**[Plant Operational
States]{.underline}**[]{#__RefHeading___Toc193082684 .anchor}

For PWRs, there are three POSs

POS 1 - This POS starts when the RHR system is placed in service. The
RCS is closed such that a steam generator could be used for decay heat
removal, if the secondary side of a steam generator is filled. 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
midloop operations with a vented RCS are subsets of this POS. Note:
performance deficiencies occurring during a 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.

For BWRs, there are three POSs

POS 1 - This POS starts when the RHR system is put into service. The
vessel head is on. This POS covers part of Hot Shutdown (Mode 3) and
Cold Shutdown (Mode 4) of the TS Modes.

POS 2 - This POS represents the shutdown condition when 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. This
POS occurs during Mode 5.

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.

**[Time Windows]{.underline}**[]{#__RefHeading___Toc193082685 .anchor}

For both BWRs and PWRs, there are two 2 Time Windows

Early Time Window (TW-E)- This time widow 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
group 3. The decay heat is relatively low. The reactor is either in POS
1, 2, or 3.

Figures 1B.1 and 1B.2 show a representative BWR and PWR outage divided
into TWs and POSs.

[]{#Table_1b1 .anchor}**Figure 1B.1 POS and Time Windows for BWRs at
Shutdown**

[]{#Table_1b2 .anchor}**Figure 1B.2 POS and Time Windows for PWRs at
Shutdown**

**[Containment Status]{.underline}**[]{#__RefHeading___Toc193082686
.anchor}

Intact Containment

An intact containment is one in which, the licensee intends to: (1)
close all containment penetrations with a single barrier, or has
procedures to enable the penetrations to be closed in time to control
the release of radioactive material, and (2) maintain the containment
differential pressure capability necessary to stay intact following a
severe accident at shutdown. When the RCS is open, an intact containment
means that the licensee intends to close the containment prior to
boiling the RCS inventory. If the licensee does not intend to maintain
an intact containment, then containment is open. To be considered an
intact containment, the containment leak rate to the environment would
be less than the values provided in Table 1A.3.

De-inerted Containment

A de-inerted containment is one in which limits on the primary
containment oxygen concentration as defined in Technical Specifications
for BWR Mark I and Mark II containments are no longer maintained.

Open Containment

An open containment is one in which the licensee does not intend to
maintain an intact containment.

**Type A Findings at Shutdown - Phase 1
Screening**[]{#__RefHeading___Toc193082687 .anchor}

For both PWRs and BWRs, In TW-L, it is assumed that LERF cannot occur
due to decay of the short-lived isotopes that are principally
responsible for early health effects (mainly I and Te). Thus, no LERF
significance determination issue is applicable for TW-L. However, LERF
can potentially occur in TW-E for both PWRs and BWRs.

For both PWRs and BWRs, a finding that is associated with a core damage
scenario is considered a potential LERF scenario during the first eight
days of shutdown. The reason for this is twofold: (1) the failures of
containment function of most concern happen relatively close to the
onset of core damage, and (2) difficulty of making a case that
evacuation would have been initiated early enough to prevent a large
release. For the findings that are screened out, the risk significance
category is determined by the CDF based SDP and need not be changed due
to LERF considerations.

**Type A Findings at Shutdown - Phase 2
Assessment**[]{#__RefHeading___Toc193082688 .anchor}

As stated above, each core damage scenario occurring during the first
eight days of shutdown is considered for LERF at shutdown. The Factor
for Type A findings related to these accidents are shown in Table 5.3
for various reactor/containment type combinations. The factors are
identified according to the status of containment.

For each core damage scenario that (1) involves an open containment and
(2) the finding occurs during the first eight days of the outage, the
risk significance category of the finding determined by the CDF based
SDP should be increased by one order of magnitude.

For each core damage scenario that (1) involves an intact containment
and (2) the finding occurs during the first eight day s of the outage,
the Factor and risk significance category of the finding is different
for the various reactor and containment types as shown in Table 5.4 of
Appendix H.

For BWR Mark I and II plants, the Factor is 1.0 as these plants are
vulnerable to hydrogen combustion events since containment is de-inerted
at shutdown. Therefore, for each LERF core damage scenario, the risk
significance category of the finding determined by the CDF based SDP
should be increased by one order of magnitude.

BWR with Mark III containments are also vulnerable to hydrogen
combustion events especially since there are no requirements for the
igniter system to be available at shutdown. If the igniters are not
available, the LERF Factor is 0.2 and the finding becomes a candidate
for increase of risk significance category. The factor of 0.2 is derived
from the discussion in Section 5.2.3 of NUREG-1765. If the igniters can
be recovered by operator action and are made available, the finding can
be screened out.

For PWR ice condenser plants, with an intact containment, there are no
technical specifications for the hydrogen igniter system to be available
once shutdown is entered. If the igniters are not available, it is
assumed that the containment is vulnerable to hydrogen combustion events
and the LERF Factor is 1.0 for the risk significant core damage
accidents. In this case, the CDF based risk significance category should
be increased by one order of magnitude. However, if the igniters can be
recovered by operator action and are made available, the finding can be
screened out.

**Type B Findings at Shutdown - Phase 1**[]{#__RefHeading___Toc193082689
.anchor}

Phase 1 screening of Type B findings at shutdown is shown in Table 6.3.
The most important feature at shutdown relative to containment SSCs for
all reactor/containment type combinations is the status of the
containment.

For PWRs and BWRs, the analyst needs to know if the containment is
intact or open. If the licensee does not intend to maintain an intact
containment, then containment is open. A Type B finding results when a
licensee intends to have an intact containment, but cannot maintain an
intact containment due to a performance deficiency.

For BWR Mark 1 and Mark II containments, it is assumed that, the
containment is de-inerted. Once POS-1 has been entered, and that a core
damage accident will lead to loss of containment function with a
conditional containment failure probability of 1.0 (due to hydrogen
combustion events). Hence containment SSCs do not play a mitigating role
in Mark I and II plants at shutdown once the containment is de-inerted,
and there are consequently no Type B findings important to ΔLERF in
these plants at shutdown.

For BWR Mark III containments and PWR ice condenser containment, the
analyst needs to know the status of the hydrogen igniters.

**Type B Findings at Shutdown - Phase 2**[]{#__RefHeading___Toc193082690
.anchor}

Phase 2 screening of Type B findings adopts generic baseline CDFs at
shutdown during POS 1E and POS 2E. These generic shutdown CDFs were
developed to support a quantitative regulatory analysis for the proposed
Shutdown Rule described in SECY 97-168. These generic CDFs assumed a
35-day refueling outage each 18 months of operation. These generic
shutdown CDFs capture cold shutdown and refueling modes until the
refueling cavity is flooded. For PWRs, the shutdown CDFs cover eight
days of cold shutdown and refueling operation until the cavity is
flooded . Of these eight days, the first two days are with the RCS
closed (POS 1); the last six days are with the RCS "vented" (POS 2). For
the BWRs, the analyses covered shutdown operation with the reactor
vessel head on for four days (POS 1) and two days with the reactor
vessel head off (POS 2).

For both PWRs and BWRs, two voluntary action cases were performed using
different interpretations of NUMARC 91-06 and GL 88-17. (NUMARC 91-06
provides guidance on improving outage management and GL 88-17 provides
recommendations concerning the ability of a licensee to mitigate a
potential loss of DHR during reduced inventory operations at PWRs.) The
higher CDF voluntary case represents a minimum implementation of both
guidance documents. The lower CDF voluntary case represents an in-depth
implementation of both guidance documents.

Based on staff review of the voluntary action cases, the staff reported
in SECY 97-168, that: (1) the existing level of safety at shutdown is
largely dependent upon measure that are not traceable to specific
underlying regulations, and that could, therefore, be withdrawn by
licensee without prior staff approval, and (2) little reduction of risk
is achieved by the rule for the licensee who has adopted effective
voluntary practices that reduced the for shutdown operation.

A licensee's LERF frequency is strongly dependent on the robustness of a
licensee's shutdown mitigation capability. A licensee's shutdown
mitigation capability is the set of systems and SSCs that can mitigate a
loss or interruption of the RHR function (a shutdown initiating event).
Therefore, the analyst must determine whether a licensee has an in-depth
shutdown mitigation capability or a minimal shutdown mitigation
capability. Tables 6.5, 6.6, 6.7 and 6.8 of Appendix H show what set of
systems and SSCs were assumed to be available for a PWR and a BWR with
an in-depth mitigation capability and with a minimal shutdown mitigation
capability.

To obtain the results in Table 6.4 of Appendix H, the high and low
voluntary action case shutdown CDFs from the Regulatory Analysis for
SECY 97-168 was used. (See Table B3). For Phase 2 analysis, each
shutdown core damage scenario was assumed to lead to a LERF scenarios
if: (1) the licensee planned but failed to maintain an intact
containment or (2) the containment fails due to loss of hydrogen control
in Ice Condenser and Mark III containments (assuming the igniters are
not available). To be considered an intact containment, the containment
leak rate to the environment would be less than the values provided in
Table 1A.3.

[]{#Table_B3 .anchor}Table B3: POSs Baseline Annualized CDF vs Plant
Mitigation Capability

+------------+------------------+-------------------+------------------+
| Reactor    | Shutdown         | Annualized CDF    | Annualized CDF   |
| Type       | Mitigation       |                   |                  |
|            | Capability       | POS 1E (Head On   | POS 2E (Head off |
|            |                  | for four days )   | for two days)    |
+------------+------------------+-------------------+------------------+
| BWR        | Minimal          | 4E-6              | 1E-6             |
+------------+------------------+-------------------+------------------+
| BWR        | In-depth         | 2E-7              | 4E-8             |
+------------+------------------+-------------------+------------------+
| PWR        | Minimal          | 3E-6              | 3E-5             |
+------------+------------------+-------------------+------------------+
| PWR        | In-depth         | 1E-7              | 8E-7             |
+------------+------------------+-------------------+------------------+

The risk results in Table 6.4 were obtained by using the following
relationship.

ΔLERF = Shutdown CDF (given in Table B3) x Factors given for Type B
findings in Table 5.2

NOTE: In Phase 3 analysis, if a licensee can show that failures
involving long term cooling can be eliminated from LERF because the
licensee would have evacuated given successful short term cooling, then
the color of the finding would be reduced.

When using Table 6.4, there are no duration factors associated with
findings at shutdown. As stated earlier, the generic shutdown CDFs
include the frequency and duration that POS 1 and POS2 are entered into
per calender year for both PWRs and BWRs. For BWRs, POS 1 is assumed to
last four days; POS 2 is assumed to last two days. For PWRs, POS 1 is
assumed to last two days; POS 2 is assumed to last six days. However,
should the duration of a Type B finding exist for less than eight hours,
then the color finding should be reduced by one order of magnitude as
part of the Phase 2 assessment.

END