about 2.1x10° per year when corrected for the 1990s Navy data and 100 lifts per year. This
estimate exceeds the proposed poo! performance guideline of 1x10° per year. The staff
believes that a licensee which chooses the non-single failure-proof handling system option in
NUREG-0612 can reduce this estimate to the same range as that for single failure-proof
systems by performing a comprehensive and rigorous load drop analysis. The load drop
analysis is assumed to include implementation of plant modifications or load path changes to
assure the SFP would not be catastrophically damaged by a heavy load drop.

6. REFERENCES

(1) U.S. Nuclear Regulatory, "Control of Heavy Loads at Nuclear Power Plants, Resolution of
Generic Technical Activity A-36," NUREG-0612, July 1980.

(2) Pittsburgh, Westinghouse, P.A., and Carlsbad, WID, N.M., "Waste Isolation Pilot Plant
Trudock Crane System Analysis," WIPP/WID-96-2196, October 1996.

(3) Richard Dudley, NRC memorandum to Document Control Desk, “Transmittal of
Information Received From the Nuclear Energy Institute (NEI) For Placement In The
Public Document Room,” dated September 2, 1999.

(4) Swain, A.D., and H.E. Guttmann, “Handbook of Reliability Analysis with Emphasis on
Nuclear Power Plant Applications,” NUREG/CR-1278, August 1983.

(5) P.G. Prassinos, et al., "Seismic Failure and Cask Drop Analyses of Spent Fuel Pools at
Two Representative Nuclear Power Plants," NUREG/CR-5176, LLNL, January 1989.

Appendix 2C. A2C-4


On high ruthenium source term, The staff concludes:

.

For the first 1 to 2 years following shutdown, the early fatality risk for an SFP fire is low, but
may be comparable to that for a severe accident in an operating reactor (based on the two
operating reactors considered). At 5 years following shutdown, the early fatality risk for
SFP accidents is approximately two orders of magnitude lower than at shutdown. This is
attributable to the effect of ruthenium, which decays to negligible amounts at 5 years.

Societal risk for an SFP fire may be comparable to that for a severe accident in an
operating reactor, but does not exhibit a substantial reduction with time because of the
slower decay of fission products and the interdiction modeling assumptions that drive long-
term doses.

Of the SFP accidents assessed, only the cask drop accident is affected by changes to EP
requirements. However, these changes do not substantially impact the total risk because
the frequency of cask drop accidents is very low. As discussed previously, changes to EP
requirements affect only the risk from cask drop accidents in the time period between 1 and
5 years.

These observations are valid regardless of whether seismic event frequencies are based
on the LLNL or the EPRI seismic hazard study.

About the low ruthenium source term the staff concludes:

Use of the low ruthenium source term reduces early fatality risk by about a factor of 100
(relative to the high ruthenium source term) within the first 1 to 2 years and by about a
factor of 10 at 5 years and after.

With the low ruthenium source term, the early fatality risk for SFP accidents is about an
order of magnitude lower than the corresponding values for a reactor accident shortly
following shutdown and about two orders of magnitude lower at 2 years following shutdown.
(in making these comparisons it is important to compare the SFP risks based on a
particular seismic hazard estimate, e.g., EPRI, with reactor accident risks based on the
same hazard estimate.)

With the low ruthenium source term, the societal risk for SFP accidents is also about an
order of magnitude lower than the corresponding values for a reactor accident shortly
following shutdown, but does not exhibit a substantial reduction with time because of the
slower decay of fission products and the interdiction modeling assumptions (discussed in
Appendix 4) that drive long-term doses. Substantial reductions would only occur after
about 5 years, when sufficient time appears to be available to initiate unplanned accident
management recovery actions.

As with the high ruthenium source term, changes to EP requirements affect the cask drop

accident, and do not substantially impact the total risk due to the low frequency of cask drop
accidents.

3-45


Introduction

The staff has performed a series of calculations to assess the potential for a criticality accident
in the spent fuel pool of a decommissioned nuclear power plant. This work was undertaken to
support the staffs efforts to develop a decommissioning rule. Unlike operating spent fuel
storage pools, decommissioned pools will have to store some number of spent fuel assemblies
which have not achieved full burnup potential for extended periods of time which were used in
the final operating cycle of the reactor. These assemblies constitute approximately one third of
the assemblies in the final operating cycle of the reactor. These assemblies are more reactive
than those assemblies normally stored in the pool which have undergone full burnup. Operating
reactors typically only store similarly reactive assemblies for short periods of time during
refueling or maintenance outages. As we will see in this report, the loss of geometry alone
could cause a criticality accident unless some mitigative measures are in place.

When spent fuel pools were originally conceived, they were intended to provide short term
storage for a relatively small number of assemblies while they decayed for a period of time
sufficient to allow their transport to a long term storage facility. Because a long term storage
facility is not available, many reactor owners have had to change the configuration of their
spent fuel pools on one or, in some cases, several occasions. This practice has led toa
situation where there are many different storage configurations at U.S. plants utilizing some
combination of geometry, burnup, fixed poisons, and boration, to safely store spent fuel.

The current state of spent fuel pools significantly complicates the task of generically analyzing
potential spent fuel pool storage configurations. Therefore, the staff decided to take a more
phenomenalogical approach to the analysis. Rather than trying to develop specific scenarios
for the different types of loading configurations, we decided to analyze storage rack
deformation and degradation by performing bounding analyses using typical storage racks.
The results of these analyses will be used to formulate a set of generic conclusions regarding
the physical controls necessary to prevent criticality. The impact of five pool storage
assumptions on the conclusions in this report will be discussed throughout the text.
Furthermore, for the purposes of this work, it is assumed that the postulated criticality event is
unrecoverable when the water level reaches the top of the fuel. This means that events such
as a loss of water leading to a low density optimal moderation condition caused by firefighting
equipment will not be considered.

It is important to reinforce the point that these analyses are intended as a guide only and will be
used to evaluate those controls that are either currently in place or will need to be added to
maintain subcriticality. These analyses will not be used to develop specific numerical limits
which must be in place to control criticality as they cannot consider all of the possible plant
specific variables. We will, however, define the controls that would be effective either
individually or in combination to preclude a criticality accident.

Description Of Methods
The criticality analyses were performed with three-dimensional Monte Carlo methods using
ENDF/B-V based problem specific cross sections (Ref. 1). Isotopic inventories were predicted

using both one- and two-dimensional transport theory based methods with point depletion.
SCALE 4.3 (Ref. 2) was used to perform the Monte Carlo, one-dimensional transport, cross

Appendix 3 A3-5


4.1.4.3 Quantification

Human Error Probabilities

The probability of failure to repair SFPC system is represented by the exponential repair model:

ew

where
A = (inverse of mean time to repair)
t = available time

In the case where discovery was from the control room, probability of failure to repair SFPC
system event, HEP-COOL-REP-E, would be 0.18 based on 17 hours available to repair.

In the case that the discovery was because of operator walk-down (HEP-COOL-REP-L), it is
assumed that there is not enough time available to repair and restart the SFP makeup system in
time to prevent bulk boiling, and the event has been assigned a value of 1.0.

4.1.4.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-COOL-REP-E 1.8E-1
HEP-COOL-REP-L 1.0

4.1.5 Top Event OFD — Operator Recovery Using Onsite Sources

4.1.5.1 Event Description and Timing

On the two upper branches of the event tree, the operators have recognized the loss of the
SFPC system, and have tried unsuccessfully to restore the system. After 43 hours, the level of
the pool has dropped below the suction of the SFP cooling system (see below), so that repair of
that system will not have any effect until pool level is restored. The operating staff now has

88 hours to provide make-up to the pool using firewater (or other available onsite sources) to
prevent fuel uncovery (131 hours less 43 hours). This event represents failure to provide make-
up to the SFP. The operators have both an electric and a diesel-driven firewater pump available
to perform this function. If both pumps were to fail, there may be time to repair one of the
pumps. This event has been modeled by the fault tree LOC-OFD.

Given the operators were not successful in detecting the loss of cooling early enough to allow
recovery of the normal cooling system, this event is modeled by functional fault tree
LOC-OFD-L. At this stage, even though the operators have failed over several shifts to detect
the need to respond, there would be several increasingly compelling cues available to the
operators performing walk-downs, including a visibly lowered pool level and a hot and humid
atmosphere. Since there are on the order of 88 hours before the level drops to 3 feet above the
fuel, some credit has been taken for subsequent crews to recognize the loss of cooling and take
corrective action.

Appendix 2A A2A-19


3.3.2 Important Assumptions

As discussed in more detail in Appendix 2A, the results of the risk analysis depend on
assumptions about the design and operational characteristics of the SFP facility. The following
inputs can significantly influence the results:

The modeled system configuration is described in Section 3.2. The assumed availability of
a diesel-driven fire pump is an important factor in the conclusion that fuel uncovery
frequency is low for the loss of offsite power initiating events and the internal fire initiating
event. The assumption of the availability of a redundant fuel pool cooling pump is not as
important since the modeling of the recovery of the failed system includes repair of the
failed pump as-well as the startup of the redundant pump. Finally, multiple sources of
makeup water are assumed for the fire pumps. This lessens the possible dependencies
between initiating events (e.g., severe weather, high winds, or earthquakes) and the
availability of makeup water supply (e.g., the fragility of the fire water supply tank).

Plants have no drain paths in their SFPs that could lower the pool level (by draining,
suction, or pumping) more that 15 feet below the normal pool operating level, and licensees
must initiate recovery using offsite sources.

Openings in the SFP building (e.g., doors and roof hatches) are large enough that, if forced
circulation is lost, natural circulation cooling will provide at least two building volumes of air
per hour to the SFP. Procedures exist to implement natural circulation.

Credit is taken for the industry/NE] commitments as described in Section 3.2. Without this
credit, the risk is estimated to be more than an order of magnitude higher. Specifically —

- IDC #1 is credited for lowering the risk from cask drop accidents.

— IDCs #2, 3, 4, and 8 are credited for the high probability of recovery from loss of cooling
scenarios (including events initiated by loss of power or fire) and loss of inventory
scenarios. To take full credit for these commitments, additional assumptions have been
made about how these commitments will be implemented. Procedures and training are
assume to give explicit guidance on the capability of the fuel poo! makeup system and on
when it becomes essential to supplement with alternate higher volume sources.
Procedures and training are assumed to give sufficiently clear guidance on early
preparation for using the alternate makeup sources. Walkdowns are assumed once per
shift and the fuel handlers are assumed to document their observations in a log. The last
assumption compensates for potential failures of the instrumentation monitoring the
status of the pool.

- IDC #5 is credited for the high probability of early identification and diagnosis (from the
control room) of loss of cooling or loss of inventory.

~ IDCs #6, 7, and 9 are credited with lowering the initiating event frequency for the loss of

inventory event from historical levels. In addition, these commitments are used to justify
the assumption that a large noncatastrophic leak rate is limited to approximately 60 gom

3-11


amplification of ground motion. In addition, there have been numerous seismic reflection and
refraction studies of the site areas for the site evaluations, and for petroleum exploration and
geophysical research. They, along with other well-proven methods, were used to determine
the nature of the geologic structure in the site vicinity, the location of any faults, and the nature
of the faults. None of these studies have indicated anomalous conditions, like those postulated
for Santa Monica, at either SONGS or DCPP. In addition, the empirical ground motion
database used to develop the ground motion attenuation relationships contains events
recorded at sites with anomalous, as well as typical ground motion amplitudes. The design
basis ground motion for both SONGS and DCPP were compared to 84th percentile level of
ground motion obtained using the attenuation relationships and the appropriate earthquake
magnitude, distance and geology for each site. The geology of the SONGS and DCPP sites
does not cause anomalous amplification, therefore, the information gained from the Kobe and
Northridge events does not raise safety concerns for U.S. nuclear power plants.

In summary, earthquakes of the type that occurred in Kobe and Northridge are different from
those that can occur near nuclear power plants in the central and eastern U.S. The higher
ground motions recorded in the Santa Monica area from the Northridge earthquake were due to
the specific geology through which the waves traveled. Improvements in our understanding of
central and eastern U.S. geology, seismic wave attenuation, seismicity, and seismic hazard
calculation methodology result in less uncertainty and lower hazard estimates today than have
previous studies.

Comment #58: The use of Lawrence Livermore National Laboratory (LLNL) hazard curves at
high ground motion values may not be credible. Even EPRI results are likely to be overly
conservative at high ground motions. The requirement that some plants with higher SSE
values perform detailed HCLPF assessments of their SFPs is not warranted. In conclusion,
there should be no SFP screening level distinctions based on plant SSEs for the central and
eastern U.S. All that is needed is that the sites pass the screening criteria (Appendix 2b). For
a few western sites, it is reasonable to require that the plants demonstrate a HCLPF of 2 X
SSE?

Response: The staff is no longer using the 2 or 3 times the SSE as a criterion. The staff
agrees that there is considerable uncertainty in the EPRI and LLNL hazard curves at higher
ground motions, since the geologic record east of the Rocky Mountains is sparse and does not
provide many examples of very large ground motions. The EPRI and LLNL hazard curves
were developed as best estimates and were made by different experts who gave their best
judgement as to how to reflect the risks from seismic events at various nuclear power plant
sites. They provided expert advice for high and low ground motions. The staffs re-evaluation
is discussed in this report.

Comments #59 and 60: (A) The NRC should determine the qualifications and degradation of
spent fuel racks. (B) How can there be no spent fuel poo! degradation issues if type 304
stainless steel employed in fuel racks and assemblies is known to exhibit stress-corrosion
cracking in oxygenated or stagnant borated water?

Response: Spent fuel rack designs do have qualifications. The designs are reviewed and
approved by the NRC. Additionally, when a licensee changes its technical specification for the
amount of fuel allowed to be stored in the pool even using approved spent fuel racks, an NRC
review and approval is required. The staff technical reviewers use the guidelines in NUREG-
0800, Standard Review Plan (SRP), which incorporates the regulations specified in the Code of

Appendix 6 A6-20


. Procedure explicitly states that if the water level drops below a certain level (e.g., 15 ft
below normal level) operator must initiate recovery using offsite sources.

° — Offsite resources are familiar with the facility.

4.4.5.3 Quantification

Human Error Probability

The event HEP-INV-OFFST-SW represents failure to take the extreme measure of using offsite
sources, given that even though there has been ample time up to this point to attempt recovery
of both the SFP cooling system and both firewater pumps it has not been successful. This top
event includes failures of both the diagnosis of the need to provide inventory from offsite
sources, and the action itself. The contribution from the failure to diagnose is assessed by
assuming a low level of dependence to account for the possible detrimental effects of the failure
to complete prior tasks successfully. A relatively low contribution of 3E-02 is assumed for failure
to complete the task, based on the fact that there are between five and six days for recovery of
the infrastructure following a severe weather event. This results in a total HEP of 8E-02. NEI
commitments 3 and 4 provide a basis for this relatively low number.

4.4.5.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-INV-OFFST-SW 8.0E-2

44.6 Summary

Table 4.4 presents a summary of basic events used in the event tree for Loss of Offsite Power
from severe weather events.

As in the case of the loss of offsite power from plant centered and grid related events, based on
the assumptions made, the frequency of fuel uncovery can be seen to be very low. Again, a
careful and thorough adherence to NEI commitments 2, 5, 8 and 10, the assumption that walk-
downs are performed on a regular, (once per shift) basis is important to compensate for
potential failures to the instrumentation monitoring the status of the pool, the assumption that the
procedures and/or training are explicit in giving guidance on the capability of the fuel pool
makeup system, and when it becomes essential to supplement with alternate higher volume
sources, the assumption that the procedures and training are sufficiently clear in giving
guidance on early preparation for using the alternate makeup sources, are crucial to
establishing the low frequency. NEI commitment 3, related to establishing communication
between onsite and offsite organizations during severe weather, is also important, though its
importance is somewhat obscured by the assumption of dependence between the events OMK
and OFD. However, if no such provision were made, the availability of offsite resources could
become more limiting.

Appendix 2A A2A-50


Figure 4.4 Severe weather related loss of offsite power event tree

[isso cerare] ovraverower] cosine | oreravor | _RecovERY ] |
Fearn! Tacsee"| Geom | seem | “a | |
[sevERE WEATHER PRIOR To sec | RESTART USING crrsire | ' |
Pievents | systeutoss | Ano RUN |waxcursystem| SOURCES | | i
t i
} 1etp2 | PR ocs ‘OMK oo | # | SEQUENCE NAMES END-STATENAMES FREQUENCY
I H i |
| 7
| if
i
i H
| | |
| 4 | IELP2 OK i
! |
| }
H i
2 |eLp20cs [9K |
i i i
recs | : 1
3 |IELP20csomK =| OK |
wacms ' : |
i
eu | i |
Keer ——— 4 IE-LP20CSOMKOFD | SFrerr | 8.425E-009
i i
| | a
5 | IELPZOPR lox i |
|
won | | |
6 | 1E-LP20PROMK OK i
LPz0Mke. | |
|
H |
L720 7 | IE-LP20PROMKOFD | serorr | o.66ze-008
i

4.4.3.2 Relevant Assumptions
. The operators have 9 hours to start the SFP cooling system before boil-off starts.

. Operators have received formal training and there are procedures to guide them (NEI
commitment no. 2).

4.4.3.3 Quantification
Human Error Probabilities

HEP-DIAG-SFPLP2 represents failure of the operator to recognize the loss of SFP cooling.
Success could result from recognition that the electric pumps stop running once power is lost
and require restart following recovery of power. If the operator fails to make an early diagnosis
of loss of SFP cooling, then success could still be achieved during walk-downs following the loss
of offsite power. Alternatively, if power is restored, the operator will have alarms available as

Appendix 2A A2A-45


AVAILABILITY OF REFERENCE MATERIALS
IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access
NUREG-series publications and other NRC records at
NRC's Public Electronic Reading Room at
www.nrc.gov/NRC/ADAMS/index.html.

Publicly released records include, to name a few,
NUREG-series publications; Federal Register notices;
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reports; and Commission papers and their
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NRC publications in the NUREG series, NRC
regulations, and Title 10, Energy, in the Code of
Federal! Regulations may also be purchased from one
of these two sources.
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Legaily binding regulatory requirements are stated
only in laws; NRC regulations; licenses, including
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NUREG-series publications. The views expressed
in contractor-prepared publications in this series
are not necessarily those of the NRC.

The NUREG series comprises (1) technical and
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(NUREG/CR-XXxXxX), (2) proceedings of
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resulting from international agreements
(NUREG/IA-XXXxX), (4) brochures
(NUREG/BR-XXxXxX), and (5) compilations of legal
decisions and orders of the Commission and
Atomic and Safety Licensing Boards and of
Directors’ decisions under Section 2.206 of NRC's
regulations (NUREG-0750).



. Operators perform walk-downs and control room readouts once per shift (every 8 to
12 hours) and document observations in a log.

. Regular test and maintenance is performed on instrumentation (NEI commitment no. 10).

. During walk-downs, level changes in the SFP can be observed on a large, graduated level
indicator in the pool.

. Procedures are available to guide the operators on response to off-normal conditions, and
the operators are trained on the use of these procedures (NEI commitment no. 2)

Appendix 2A A2A-15


measures is also substantially lower than that allowed in RG 1.174. Since the SFP fire
frequency assumed in these comparisons is about a factor of 4 lower than the PPG of 1x10°
per year, an SFP facility operating nominally at the PPG would have a smaller margin to the
allowable risk limits for the reference plant but would still be at or below these limits.

As discussed in Section 3.7, the basis for these results is that EP is of marginal benefit in large
earthquakes because of offsite damage. However, as described in Appendix 4D, even with
unrealistically optimistic assumptions about the effectiveness of EP in seismic events (i.e.,
assuming full and relaxed EP results in early and late evacuation, respectively, and using the
LLNL seismic hazard frequency and the high ruthenium source term), the change in risk is small
and the QHOs continue to be met with adequate margin.

Measures to Assure Risk Increases Remain Small

The analysis in Section 3 explicitly examines the risk impact of specific design and operational
characteristics. This analysis credits the industry decommissioning commitments (IDCs)
proposed by NEI in a letter to the NRC dated November 12, 1999 (Ref. 4) and several additional
staff decommissioning assumptions (SDAs) identified through the staff's risk assessment and
the staffs evaluation of the RG 1.174 safety principles for decommissioning plants. The IDCs
and SDAs are summarized in Tables 4.2-1 and 4.2-2.

The low numerical risk results shown in Section 3 and Appendix 2 are predicated on the IDCs
and SDAs being fulfilled. Specifically,

+ IDC #5 and SDAs #2 and #3 provide assurance of timely operator response for a broad
range of operational events.

* The low likelihood of pool failure due to heavy load drop is dependent on design and
procedural controls for handling of heavy loads (IDC #1 and #9 and SDA #5).

+ The low baseline frequency for a seismically initiated zirconium fire is predicated upon
implementation of the seismic checklist shown in Appendix 5 (SDA #6).

+ — The low likelihood of loss of cooling is dependent upon procedures and training (IDC #2)
and instrumentation (IDC #5 and SDA #3).

+ The low likelihood of loss of inventory is dependent upon design provisions (IDC #6) and
procedures and controls (IDC #7) to limit leakage.

+ The high probability that the operators will identify and recover from a loss of cooling or a
loss of inventory event is dependent upon procedures and training for effective use of
onsite and offsite resources (IDCs #2 through #4, IDC #8, and SDA #3) and SFP
instrumentation (IDC #5 and SDA #3).

+ The low likelihood criticality issues is dependent on continuation of programs to assess the
condition of Boraflex absorber material (SDA #7).

45


Figure A2c-1 (sheet 1 of 2) - Heavy Load Drop Fault Trees

FHLS

FAILURE OF HEAVY LOAD
(CRANE AND RIGGING) SYSTEM

CRANE RIGGING
FAILURE FAILURE
| OF
CRANE RIGGING
cer - 1 cF2 CF4 CF3 CR11 cri2
Failure due Failure of 7 ‘
Load "Hangup” to Random Component ‘Two-blocking improper eee oe
Event Component Without backup Event aor 8 Rigging
Failure (NO"Fa"Fa")
Crit cri2 cr21_ | |__cr22 ors: [ . cF32 ]_crss
‘Operator error Failure of single 7 ‘Operator error i
Leading to load Fallure of Component with ff Failure of aclup Leading ta Failure of ohper
angup a bac! “Two-blockin,
(NOTE) Device {NO"F2) Primary component (NOES) iF |] Limit switch |] Limit switch

A2C-9


the SPAR HRA technique. The assumptions included expansive time (> 24 hours), extreme
stress, highly trained staff, diagnostic type procedures, and good quality of work process. This
diagnosis task provides the diagnosis for the subsequent actions taken to re-establish cooling to
the pool.

HEP-FW-START-SW represents failure to start either the electric or diesel firewater pump
(depending upon availability) within 88 hours after the onset of bulk boiling, given that the
decision to start a firewater pump was made. No difficult valve alignment is required, but the
operator may have to run the fire hoses to designated valve stations. This event was quantified
using the SPAR HRA technique. The PSFs chosen were; expansive time (> 50 times the
required time), high stress, highly complex task because of the multiple steps and severe
weather and its non-routine nature, quality procedures, poor ergonomics because of severe
weather, and finally a crew who had executed these tasks before, conversant with the
procedures and one another.

HEP-FW-REP-NODSW represents the failure of the repair crew to repair a firewater pump for
the scenario where power is not recovered. Note that we have assumed that since power is not
recovered, the repair crew did not make any attempt to repair the SFPC system, and therefore
no dependency was modeled in the failure to repair the firewater system. We assume that the
operator will focus his recovery efforts on only one pump. Assuming that it takes two days

(48 hours) before technical help and parts arrive, then the operator has 40 hours (88 hours less
48 hours) to repair the pump. Assuming a 10-hour mean time to repair, the probability of failure
to repair the pump would be Exp [-(1/10) ( 40)] = 1.8E-2. This event is modeled in the fault tree,
LP2-OMK-L.

HEP-FW-REP-DEPSW represents the failure of the repair crew to repair a firewater pump for
the scenario where power is recovered. Note that repair was not credited for top event OCS;
however, we have assumed that the repair crew did make an attempt to restore the SFPC
system, and so dependency was modeled in the failure to repair the firewater system. For
HEP-FW-REP-DEPSW a low level of dependence was applied modifying the failure rate of
2.5E-2 to 7.0E-2 using the THERP formulation for low dependence.

In addition, in fault tree LP2-OMK-U, the possibility that no action is taken has been included by
incorporating an OR gate with basic events HEP-DIAG-SFPLP2 and HEP-RECG-DEPEN. The
latter is quantified on the assumption of a low dependency.

Non-HEP Probabilities

In the case of LP2-OMK-U, both firewater pumps are available. Failure of both firewater pumps
is represented by basic event FP-2PUMPS-FTF.

In the case of LP2-OMK-L, only the diesel-driven firewater pump is available, and its failure is
represented by basic event FP-DGPUMP-FTF.

The pump may be required to run 8 to 10 hours at the most (250 gpm capacity), given that the
water inventory drops by 20 ft (i.e., 3 ft above the top of the fuel). A failure probability of 3.7E-3
for failure to start and run for the electric pump and 0.18 for the diesel driven pump are used

Appendix 2A A2A-48


effort along with emergency preparedness and security requirements. In the past, licensees
have been granted exemptions from financial protection requirements on the basis of
deterministic analyses that indicate that a zirconium fire could no longer occur.

In the staff requirements memorandum (SRM) for SECY-93-127 (Ref. 11), the Commission
suggested that withdrawal of secondary financial protection insurance coverage be allowed after
the requisite minimum spent fuel cooling period had elapsed. Further, the Commission directed
the staff to determine more precisely the appropriate spent fuel cooling period after plant
shutdown. While insurance does not lend itself to a “small change in risk” analysis because
insurance affects neither the probability nor the consequences of an event, the NRC staff has
considered whether the risk analysis in this study justifies relief from this requirement for a
decommissioning plant while it is vulnerable to zirconium fires. The risk analysis in the February
2000 study identified a generic window of vulnerability for an SFP fire until about 5 years after
shutdown. The analysis in this study, however, indicates that a zirconium fire cannot be
precluded on a generic basis even after 5 years decay. This is because a spent fuel
configuration necessary to assure air cooling cannot be assured following a severe earthquake
or cask drop event that drains the pool. Since a criteria of “sufficient cooling to preclude a fire”
cannot be met and the long-term consequences could be significant (e.g., the long-term
consequences (and risk) decrease very slowly because cesium-137 has a half life of
approximately 30 years), the staff will need to consider alternative criteria if changes to
insurance requirements are to be pursued.

4-16


PWR Low Density Storage Rack Eigenvalue Following Compressive/Expansion

Figure 2

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A3-15

Appendix 3


3.0 RISK ASSESSMENT OF SPENT FUEL POOLS AT DECOMMISSIONING PLANTS

The scenarios leading to significant offsite consequences at a decommissioning plant are
different than at an operating plant. Once fuel is permanently removed from the reactor vessel,
the primary public risk in a decommissioning facility is associated with the spent fuel pool (SFP).
The spent fuel assemblies are retained in the SFP and submerged in water to cool the
remaining decay heat and to shield the radioactive assemblies. The most severe accidents
postulated for SFPs are associated with the loss of water from the pool.

Depending on the time since reactor shutdown, fuel burnup, and fuel rack configuration, there
may be sufficient decay heat for the fuel clad to heat up, swell, and burst after a loss of pool
water. The breach in the clad releases of radioactive gases present in the gap between the fuel
and clad. This is called “a gap release” (see Appendix 1B). If the fuel continues to heat up, the
zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or
zirconium and steam is exothermic (i.e., produces heat). The energy released from the reaction,
combined with the fuel’s decay energy, can cause the reaction to become self-sustaining and
ignite the zirconium. The increase in heat from the oxidation reaction can also raise the
temperature in adjacent fuel assemblies and propagate the oxidation reaction. The zirconium
fire would result in a significant release of the spent fuel fission products which would be
dispersed from the reactor site in the thermal plume from the zirconium fire. Consequence
assessments (Appendix 4) have shown that a zirconium fire could have significant latent health
effects and resulted in a number of early fatalities. Gap releases from fuel from a reactor that
has been shutdown more than a few months involve smaller quantities of radionuclides and, in
the absence of a zirconium fire, would only be of concern onsite.

The staff conducted its risk evaluation to estimate the likelihood of accident scenarios that could
result in loss of pool water and fuel heatup to the point of rapid oxidation. In addition to
developing an assessment of the level of risk associated with SFPs at decommissioning plants,
the staff's objective was to identify potential vulnerabilities and design and operational
characteristics that would minimize these vulnerabilities. Finally, the staff assessed the effect of
offsite emergency planning (i.e., evacuation) at selected sites using various risk metrics and the
Commission's Safety Goals.

In support of the risk evaluation, the staff conducted a thermal-hydraulic assessment of the SFP
for various scenarios involving loss of pool cooling and loss of inventory. These calculations
provided information on heatup and boiloff rates for the pool and on heatup rates for the
uncovered fuel assemblies and time to initiation of a zirconium fire (see Table 2.1 and

Figures 2.1 and 2.2). The results of these calculations provided fundamental information on the
timing of accident sequences, insights on the time available to recover from events, and time
available to initiate offsite measures. This information was used in the risk assessment to
support the human reliability analysis of the likelihood of refilling the SFP or cooling the fuel
before a zirconium fire occurs.

For these calculations, the end state assumed for the accident sequences was the state at
which the water level reached 3 feet from the top of the spent fuel. This simplification was used
because of the lack of data and difficulty in modeling complex heat transfer mechanisms and
chemical reactions in the fuel assemblies that are slowly being uncovered. As a result, the time

3-1


Attachment 2C-1
Uncertainties
Incident rate.

The range used in this evaluation (1.0x10* to 1.5x10* incidents per year) was based on
the Navy data originally assessed by the staff in NUREG-0612. The 1999 Navy data, like
the 1980 data, did not report the number of lifts made and only provided information about
the number of incidents. The cask loading experience at light water reactors and Ft. St.
Vrain tends to support values used for the incident range.

Drop rate.

The drop rate, about 1-in-10, was based on the 1999 Navy data. Previous studies used
engineering judgement to estimate the drop rate to be as low as 1-in-100.

Load path.

The fraction of the load path over which a load drop may cause sufficient damage to the
SFP to result in a loss-of-inventory was estimated to be between 0.5 percent and

6.25 percent of the total path needed to lift, move, and set down the load. This range was
developed by the staff for the NUREG-0612 evaluation. No time motion study was
performed to account for the fraction of time the load is over any particular location.

Load handling design.

The benefit of a single-failure-proof load handing system to reduce the probability of a load
drop was estimated to be about a factor of 10 to 100 improvement over a non-single
failure-proof load handling system, based on the fault tree quantifications in this
evaluation. Previous studies have used engineering judgement to estimate the benefit to
be as high as 1,000.

Load drop analysis
The benefit of a load drop analysis is believed to be significant, but is unquantified. A load
drop analysis involves mitigation of the potential drop by methods such as changing the

safe load path, installation of impact limiters, or enhancement of the structure, as
necessary, to be able to withstand a heavy load drop at any location on a safe load path.

A2C-5


because NUREG/CR-4982 was for an operating reactor and this analysis is for a
decommissioned reactor.

Because NUREG/CR-4982 was a study of SFP risk for an operating reactor, the Millstone 1
SFP inventories shown in Table A4-1 were for the fuel that was discharged during the 11"
refueling outage (about 1/3 of the core) and the previous 10 refueling outages. The inventories
shown in Table A4-1 did not include the fuel which remained in the vessel (about 2/3 of the
core) that was used further when the reactor was restarted after the outage. Because the
current study is for a decommissioned reactor, the inventories shown in Table A4-1 were
adjusted by adding the inventories in the remaining 2/3 of the core. This remaining 2/3 of the
core is expected to contain a significant amount of short half-life radionuclides in comparison
with the 11 batches of spent fuel in the SFP.

The radionuclide inventories in the remaining 2/3 of the core were derived from the data in
Tables A.5 and A.6 in NUREG/CR-4982. Tables A.5 and A.6 give inventory data for the 11"
refueling outage. Table A.5 gives the inventories for the entire core at the time of reactor
shutdown. Table A.6 gives the inventories (at 30 days after shutdown) for the batch of fuel
discharged during the outage. First, the inventories for the entire core at the time of shutdown
were reduced by radioactive decay to give the inventories for the entire core at 30 days after
shutdown. Then, the inventories (at 30 days after shutdown) for the batch of fuel discharged
were subtracted to give the inventories for the remaining 2/3 of the core at 30 days after
shutdown. Inventories for the remaining 2/3 of the core at 90 days and 1 year after shutdown
were subsequently calculated by reducing the 30-day inventories by radioactive decay.

Appendix 4 A4-2


NUREG/CR-6451 included the population density, the amount of spent fuel in the pool, and the
exclusion area size. To provide a more consistent basis to compare the NUREG/CR-6451
results with the Case 4 results, Case 4 was rerun using population densities, an amount of
spent fuel, and an exclusion area size similar to NUREG/CR-6451.

The average population densities in the NUREG/CR-6451 analysis were about 1800
persons/mile? within 50 miles and 215 persons/mile? within 500 miles. Also, NUREG/CR-6451
used an inventory with substantially higher quantities of long-lived radionuclides than the 11
batches of spent fuel in NUREG/CR-4982. NUREG/CR-6451 stated that it used an inventory of
Cs-137 (30 year half-life) that was three times greater than that used in NUREG/CR-4982. To
provide a more consistent basis to compare with NUREG/CR-6451 long-term consequences,
Case 4 was rerun using uniform population densities of 1800 persons/mile? within 50 miles and
215 persons/mile? outside of 50 miles and a power correction factor of 3 instead of 1.7. As
shown in Table A4-15, Case 4 rerun is in generally good agreement with NUREG/CR-6451.
These calculations indicate a very strong dependence of long-term consequences on
population density. Remaining differences in long-term consequences may be because of
remaining differences in population density and inventories as well as differences in
meteorology and emergency response.

Table A4-15 Comparison with NUREG/CR-6451 Results (long-term consequences)

Dist. Societal Dose (person-Sv) Cancer Fatalities
(miles) TNUREG | Case 4 Case 4 NUREG/ | Case4 | Case 4
CR-6451 Rerun CR-6451 Rerun
0-50 750,000 23,600 389,000 31,900 1,260 20,800
0-500 3,270,000 454,000 1,330,000 138,000 20,600 44,900

To provide a more consistent basis to compare with NUREG/CR-6451 short-term
consequences, Case 4 was again rerun, this time using a uniform population density of 1000
persons/mile? and an exclusion area of .32 miles. As shown in Table A4-16, Case 4 rerun is in
generally good agreement with NUREG/CR-6451. Overall, these calculations indicate a very
strong dependence of short-term consequences on population density and a small dependence
(about 10% change in prompt fatality results) on exclusion area size. Remaining differences in
short-term consequences may be because of remaining differences in population density and
inventories as well as differences in meteorology and emergency response.

Appendix 4 A4-15


SPAR HRA Human Error Worksheet (Page 2 of 3)

Plant:, Initiating Event:_ Sequence Number: Basic Event Code:

Basic Event Context:
Basic Event Description:

Part Il. ACTION
A. Evaluate PSFs for the action portion of the task.

PSFs PSF Levels Muitiplier for If non-nominal PSF levels are selected, please note specific
Action reasons in this column

‘Available Time Tnadequate time P(failure) = 1.0
““fime availabie time
required
Nominal time
‘Fime available>50 x time cx
required
‘Stress Extreme ; 5 — VS aETEAETST—e
High” z
‘Nominal i
[Complexity 3
z
; a
Experience/Training 3.
7
Procedures

Ergonomics Missing/Misleading 30
Poor _ iv

“Nominal " i
Good O35

Fitness for Duty

‘Work Processes

B. Calculate the Action Failure Probability

(1) Ifall PSF ratings are nominal, then the Action Failure Probability = 1E-3

(2) Otherwise, Time Stress Complexity Experience/ Procedures Ergonomics Fitness Work
Training for Duty Processes
Action: 1E-3 x x x x x x x x =
Action

Failure Probability


Gary M. Holahan

2

Table 2 Results based on NUREG-1465 Source Term

Cas | Decay Mean Consequences?
e Time (Surry population, 95% evacuation)

Individual Risk of | Individual Risk Societal Early

Early Fatality of Cancer Dose (rem) Fatalities

(within 1mile) Fatality (within (within 50 (within 10

10 miles) miles) miles)
77a 30 days 1.27x10? 1.88x10? 5.58x10° 2.21
77b 90 days 9.86x10° 1.82x10? 5.43x10° 1.37
T7c 1 year 7.13x10% 1.68x10? 5.28x10° .736
77d 2 years 5.64x10% 1.58x10? 5.12x10® 481
7Te 5 years 3.18x10* 1.43x10? 4.90x10® 192
77 10 1.63x10* 4.29x10? 4.72x10° .0778
years
78a’ | 30 days 8.36x10% 9.92x10* 4.12x10® .0720
78b° | 90 days 6.83x10* 9.62x10* 4.02x10® .0461
78c° 1 year 5.44x10* 9.09x10* 3.95x10° .0301
78d° 2 years 4.41x104 8.71x10% 3.87x10° 0208
78e 5 years 2.54x10% 8.14x10% 3.77x10° .00882
78f 10 1.47x10* 7.70x10* 3.69x10° .00400
years

@Accident frequencies approximately 10°/year or less.
*Based on early evacuation.


4.2.1 Initiating Event FIR - Internal Fire
4.2.1.1 Event Description and Timing

The fire initiator includes those fires of sufficient magnitude, that if not suppressed, would cause
a loss of cooling to the SFP. This loss of cooling could either result from damage to the SFPC
system or the offsite power feeder system.

4.2.1.2 Relevant Assumptions

¢ Fire ignition frequencies from operating plants are assumed to be applicable at the SFP
facility.

. Ignition sources from welding and cutting are expected to be insignificant. The facility
configuration is expected to be stable, negating the need for modification and fabrication
work requiring welding and cutting.

4.2.1.3 Quantification

Data compiled from historical fires at nuclear power plants is summarized in the Fire-Induced
Vulnerability Evaluation (FIVE) methodology document (Ref. 15). This document identifies fire
ignition sources and associated frequencies and is segregated by plant location and ignition
type. Of the plant locations identified in the FIVE document, the intake structure was considered
to most closely approximate the conditions and equipment associated with the SFP facilities
considered in this analysis.

FIVE identifies specific frequencies associated with "electrical cabinets,” "fire pumps," and
"others" in the intake structure. In addition to these frequencies associated with specific
equipment normally located in the intake structure, ignition sources from equipment (plant-wide)
that may be located in the intake structure is also apportioned.

The largest ignition frequency contribution identified for intake structures is from fire pumps. In
the plant configuration assumed in this study, the firewater pumps are located in an unattached
structure and thus can be eliminated as ignition sources. FIVE also identifies electrical cabinets
as significant ignition sources in the intake structure with an average frequency of 2.4E-3/yr.
Because the number of electrical cabinets (breakers) in the spent fuel facility is expected to be
less than those in the typical intake structure, a scaling factor was used to estimate the electrical
cabinet contribution. Typically there are five motor-driven pumps (4 cooling pumps, 1 makeup
pump) and related support equipment associated with the SPF facility. The number of electrical
cabinets (breakers) was therefore estimated to be less than ten in a typical SFP facility. The
number of electrical cabinets in the intake structure was estimated to be 25 (engineering
judgement based on plant walk-downs). Therefore, the fire ignition frequency contribution from
electrical cabinets at the SFP facility is estimated to be (10/25)(2.4E-3/yr) = 9.6E-4/yr.

Appendix 2A A2A-24


consequences was seen.

Table 8 Results of Release Fraction Sensitivities
(99.5% evacuation, Surry Population Density)

Case | Release Fraction Mean Consequences (within 100
miles)
Cs | Ru Te Ba Sr Ce La Early Societal Cancer
Fatali- | Dose Fatalities
ties (person-
rem)
Base | 1 2x10% | .02 | .002 | .002 | 1x10* | 1x10* 1.01 4.54x10® 2,320
SS
11 1 4 02 | 002 | .002 | 1x10® | 1x10* 95.3 9.53x10° 9,150
96 1 1 102 | .01 01 .01 01 106 1.33x10" 11,700
95 78 15 .02 | .01 01 .01 01 57.0 1.17x107 10,400
94 75 75 02 | .002 | .002 | .001 004 50.2 8.35x10° 7,850
Ld dd ed
14° 41 1 02 .002 | .002 | 1x10* 1x10 132 6.75x10® 6,300
97? 1 1 02 01 01 01 01 154 8.74x10® 7,990

*Based on evacuation before release.

The evaluation documented in Appendix 4 used a conservative release fraction of one for the volatile
fission products. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, February
1995, specifies a more realistic release fraction of .75 for volatile fission products. As part of the
sensitivity of the effect of fuel fines release fraction, this more realistic release fraction was used. In
Case 95, the consequences decreased as a result of decreasing the volatile fission product release
fraction from 1 to .75. In this case, a factor-of-two decrease in the early fatalities and a small decrease in
the long-term consequences were seen.

Finally, Case 94 was run to investigate the sensitivity of the consequences to a fuel fines release fraction
intermediate between 1x10* and .01. This case used a fuel fines release fraction of .001. As a result of
decreasing the fuel fines release fraction from .01 to .001, a small decrease in the consequences was
seen.

In Case 11, evacuation begins about an hour after the fission product release begins. However,
Appendix 1 states that, after a year of decay, it will take a number of hours for the fuel with the highest
decay power density to heat up to the point of releasing fission products in the fastest progressing
accident scenarios. As a result, it is more likely to have evacuation before the release begins.
Therefore, a sensitivity calculation on fuel fines release fraction also was run using Case 14 as the
starting point; Case 14 includes evacuation three hours before the release begins. Case 97 was run with
a fuel fines release fraction of .01. As a result of increasing the fuel fines release fraction from 1x10* to
.01, a small increase in the offsite consequences was seen.

The above sensitivity calculations for fuel fines release fractions were performed with 99.5% of the
population evacuating. This translates into one person in 200 not evacuating. It has been suggested
that the percentage of the population evacuating may be smaller. Therefore, the staff performed
additional calculations with 95% of the population evacuating. This translates into one person in 20 not
evacuating. The results of these calculations are shown in Table 9. Case 45, which used a ruthenium
release fraction of one, is the shown in the second row of Table 9 and was the starting point for these
calculations. Then, Case 45a was run with a fuel fines release fraction of .01, and Case 45b was run
with a volatile fission product release fraction of .75. The same trends were seen as in the 99.5%
evacuation cases, Cases 11, 96, and 95.

-6-


firewater pump and provide make-up to the SFP. If the diesel firewater pump fails, the operators
have time to attempt repair. This event is modeled by fault tree FIR-OMK.

4.2.5.2 Relevant Assumptions

. There is a means to remotely align a makeup source to the SFP without entry to the refuel
floor, so that make-up can be provided even when the environment is uninhabitable
because of steam and/or high radiation (NEI commitment no.8).

° Inventory makeup using the firewater system is initiated by onsite operators.

. In modeling the repair of a failed firewater pump, it is assumed that it takes 16 hours to
contact maintenance personnel, make a diagnosis, and get new parts.

. Mean time to repair the firewater pump is 10 hours.

« Inventory makeup using the firewater pumps is proceduralized, and the operators are
trained in these procedures (NEI commitment no. 2).

. Firewater pumps are tested and maintained on a regular schedule (NEI commitment
no. 10).

4.2.5.3 Quantification
Human Error Probabilities
The fault trees used to quantify this top event include three human failure events.

HEP-RECG-FWSTART represents the operators' failure to recognize the loss of SFP cooling
and the need to initiate the firewater system. This event was quantified using the SPAR HRA
technique. The assumptions include expansive time (> 24 hours), a high level of stress,
diagnostic type procedures, good ergonomic interface, and good quality of work process. This
diagnosis task provides the diagnosis for the subsequent actions taken to re-establish cooling to
the pool. Although this diagnosis and subsequent actions follow a fire, no dependence between
response to the fire and subsequent actions is assumed, because of the large time lag.

HEP-FW-START represents failure to start the diesel firewater pump within 88 hours after the
onset of bulk boiling, given that the decision to start a firewater pump was made. No difficult
valve alignment is required, but the operators may have to run hoses to designated valve
stations. This event HEP-FW-START was quantified using SPAR HRA technique. The
following PSFs were assumed: expansive time (> 50 times the required time), high stress, highly
complex task because of the multiple steps, its non-routine nature, quality procedures available,
as well as good ergonomics including equipment and tools matched to procedure, and finally a
crew who had executed these tasks before, conversant with the procedures and one another.

HEP-FW-REP-NODEP represents the failure of the repair crew to repair a firewater pump. It is
assume that the operators will focus their recovery efforts on only the diesel driven pump.

Appendix 2A A2A-30


RULEMAKING & NRC PROCESS CONCERNS

Comment #89: Since more radioactive materials are being handled [during decommissioning]
than at an operating plant, and under conditions more likely to lead to inadvertent exposures,
why are licensees left without the supervision of resident inspectors, or at least radiation
protection personnel?

Response: During operation of a reactor, radioactive material is produced by neutron
absorption by various materials. These radioactive materials are handled in many ways,
including liquids contained in pipes and tanks, and radioactive solids contained in plastic bags
or specialized containers. After the reactor is shut down, no additional radioactive material is
produced and the radioactive material decay process reduces the total amount of radioactive
material over time: The handling of radioactive material after shutdown is controlled in the
same manner as before shutdown. Supervision of radioactive material handling is performed
by the licensee before and after reactor shutdown with the oversight of licensee radiation
protection personnel. Region-based NRC inspectors provide periodic verification that the
licensee is handling radioactive materials within the bounds of the current regulations. NRC.
experience over the last few years with the current region-based reactor decommissioning
inspection process has shown that the oversight process is effective in ensuring both public
health and safety and protection of plant workers.

Comment #90: A commenter stated that little of what operators or reactor inspectors have
learned is applicable to decommissioning. NRC needs personnel specifically trained in and
dedicated to decommissioning.

Response: Significant changes take place during the transition from an operating plant to a
decommissioning plant. However, many decommissioning activities are similar to activities
conducted during plant operation. For example, the complete removal of components and
systems, radiological waste shipments, fuel handling operations, and spent fuel pool system
operations and maintenance which occur during decommissioning are very similar to activities
that occurred during plant operation and refueling outages. Objectives during
decommissioning, such as, protecting the spent fuel from sabotage and maintaining the spent
fuel pool operational, were also accomplished during plant operation. The training received by
operators and inspectors associated with radiological fundamentals, system operations, etc.,
still applies during decommissioning.

Although there is not an NRC inspector on-site during all of decommissioning, as there is
during plant operation, there is a group of inspectors in each region who are specifically
assigned to oversee plants undergoing decommissioning, and who make routine visits to the
site (commensurate with the quantity and significance of the ongoing work). Each plant in
decommissioning is also assigned to a project manager located at NRC Headquarters. These
project managers are assigned to a section that is responsible only for permanently shutdown
power reactors.

Comment #91: What does "reducing unnecessary regulatory burden" mean in practice, when it
comes to emergency planning? What kind of reductions are foreseen for the following:
manpower on-site/off-site, emergency equipment, communication means, alarm means,
notification of personnel/public, EP, plans, KI [potassium iodide], EPZ [emergency planning
zone] radius?

Appendix 6 A6-30


Figure A2e-3 Sketch of Hypothetical F2 Tornado Illustrating Variations

Figure A2e-4 Tornado Risk Regionalization Scheme (from NUREG/CR-2944)

Appendix 2E A2E-10


RARER

HD) th

‘eeaonresto

gevcerne

RSE

OPERATOR FAILS TO
‘OBSERVE LOULSFPC
EVENT DURING
‘WALKDOWN:

HEP-WLKDINLOHL,


PROBABILITY AND HUMAN RELIABILITY ASSESSMENT

Comment #10: Experience at nuclear power plants demonstrates that safety problems are not
caused by workers making mistakes or by not following procedures. Problems are caused by
bad management.

Response: The staff agrees that utility safety culture and utility oversight/expectations in the
day-to-day operations of a facility are important contributors to either a well run plant or a poorly
run one. The staff decommissioning assumptions and industry commitments will help insure
that proper attention is given to spent fuel pool status, procedures are developed that guide fuel
handlers in the event of a spent fuel pool accident, communications are established between
on-site and off-site organizations, and cask drop analyses are performed or a single failure
proof crane is used for handling very heavy loads. These staff assumptions and industry
commitments are discussed in Sections 3 and 4.

Comment #11: Experience at nuclear power plants shows that multiple shifts can make the
same error and not recognize it for a long time. With watching the pool being their major
responsibility, a fuel handler’s life would be very tedious and boredom would set in. This
should result in a poorer response by the fuel handler in the event of an accident. An example
of this is the recent Browns Ferry event.

Response: The NRC, through the “Policy on Factors Causing Fatigue of Operating Personnel
at Nuclear Reactors” provides guidelines on working hours that were consistent with the
objective of ensuring that the mental alertness and decision-making abilities of plant staff were
not significantly degraded by fatigue. The staff shares the commenter’s concem that operator
boredom and their ability to maintain alertness while standing watch may contribute to fatigue-
induced impairment of personnel and thereby increase the likelihood of personnel errors. For
this study, our modeling and quantification of SFP risk includes consideration of multiple shift
turnovers and the chance that shift after shift makes the same mistake. However, for almost all
postulated SFP accidents, there is a very long time available to the fuel handlers to discover
and recover from the existence of a problem in the spent fuel pool or its support systems. The
staff believes that the commitments made by the industry and the NRC’s staff decommissioning
assumptions provide a basis for reducing the chances of multiple shift errors to the point where
they do not contribute significantly to the overall risk of spent fuel pool operation (See

Sections 3 and 4). Other accidents (i.e., seismic and heavy load drop), which progress rapidly,
are assumed to proceed independent of operator intervention once the accident has occurred
because the SFP is assumed to drain very rapidly.

Comment #12: Over time, tedious tasks will cause workers to make mistakes. The NRC
needs to address this in a conservative manner.

Response: The staff agrees that tedious tasks can increase the chances of a fuel handler
making a careless mistake. The NRC, through the “Policy on Factors Causing Fatigue of
Operating Personnel at Nuclear Reactors” provides guidelines on working hours that were
consistent with the objective of ensuring that the mental alertness and decision-making abilities
of plant staff were not significantly degraded by fatigue. However, we do not agree that fuel
handler errors need be modeled in a conservative manner when performing a probabilistic risk
assessment. It is the NRC’s policy to make its risk assessments as realistic as possible. The
staff performed the analysis for this report consistent with the agency’s policy.

Appendix 6 A6-4


FAILURE TO RECOVER
SOU
(65 HRS)
a
L 1 1
FAILS TO DIAGNOSE DINGNOSERECOVER VA
NEED TO START FW ‘OFFSITE SOURCES (85 HRS)
CEPCNCENTEVENT)
cen HEPANFOFFSITE
H
FAILURE TO RECOVER:
OFFSITE SOURCES
BL
iT
F 1
FAILS TO DIAGNOSE: OSEREGOVER VA
NEED TO START FW OFFSITE SOURCES (85 HRS)
(DEPENDENT EVENT)
HEP-RECG-FWSTART-L HEP-INV-OFFSITE

A-5


Table 4.1-1 Industry Decommissioning Commitments (IDCs)

IDC No. Industry commitments

1 Cask drop analyses will be performed or single failure-proof cranes will be in
use for handling of heavy loads (i.e., phase Ii of NUREG-0612 will be
implemented).

2 Procedures and training of personnel will be in place to ensure that onsite and

offsite resources can be brought to bear during an event.

3 Procedures will be in piace to establish communication between onsite and
offsite organizations during severe weather and seismic events.

4 An offsite resource plan will be developed which will include access to portable
pumps and emergency power to supplement onsite resources. The plan would
principally identify organizations or suppliers where offsite resources could be
obtained in a timely manner.

5 SFP instrumentation will include readouts and alarms in the control room (or
where personnel are stationed) for SFP temperature, water level, and area
radiation levels.

6 SFP seals that could cause leakage leading to fuel uncovery in the event of
seal failure shall be self limiting to leakage or otherwise engineered so that
drainage cannot occur.

7 Procedures or administrative controls to reduce the likelihood of rapid

draindown events will include (1) prohibitions on the use of pumps that lack
adequate siphon protection or (2) controls for pump suction and discharge
points. The functionality of anti-siphon devices will be periodically verified.

8 An onsite restoration plan will be in place to provide repair of the SFP cooling
systems or to provide access for makeup water to the SFP. The plan will
provide for remote alignment of the makeup source to the SFP without
requiring entry to the refuel floor.

9 Procedures will be in place to control SFP operations that have the potential to
rapidly decrease SFP inventory. These administrative controls may require
additional operations or management review, management physical presence
for designated operations or administrative limitations such as restrictions on
heavy load movements

10 Routine testing of the alternative fuel pool makeup system components will be
performed and administrative controls for equipment out of service will be
implemented to provide added assurance that the components would be
available, if needed.

4-11



event, such as a steam explosion which could damage and drain the spent fuel pool. The staff
judges the chances of such a criticality event to be well below 1 chance in 100, even given that
the transfer cask drops directly onto the assemblies. This would put the significant criticality
likelihood well below 1E-8 per year.

Deformation of the low-density BWR racks by the dropped transfer cask was shown to most
likely not result in any criticality events. However, if some mode of criticality were to be induced
by the dropped transfer cask, it would more likely be a small return to power for a very localized
region, rather than the severe response discussed in the above paragraph. This minor type of
event would have essentially no off-site (or on-site) consequences since the reaction’s heat
would be removed by localized boiling in the pool and water would provide shielding to the site
operating staff. The reaction could be terminated with relative ease by the addition of boron to
the pool. Therefore, the staff believes that qualitative, as well as some quantitative assessment
of scenario 1 demonstrates that it poses no significant risk to the public from SFP operation
during the period that the fuel remains stored in the pool.

With respect to scenario 2 from above (i.e. the gradual degradation of the Boraflex absorber
material in high-density storage racks), there is currently not sufficient data to quantify the
likelihood of criticality occurring because of its loss. However, the current programs in place at
operating plants to assess the condition of the Boraflex, and take remedial action if necessary
provide sufficient confidence that pool reactivity requirements will be satisfied. In order to meet
the RG 1.174 safety principle of maintaining sufficient safety margins, the staff judges that
continuation of such programs into the decommissioning phase should be considered at all
plants until all high-density racks are removed from the SFP.

Based upon the above conclusions and staff requirements, we believe that qualitative risk
insights demonstrate conclusively that SFP criticality poses so meaningful risk to the public.

3.2.2 Deterministic Criticality Study
This section includes a copy of the report entitled "Assessment of the Potential for Criticality in

Decommissioned Spent Fuel Pools” which is a deterministic study of the potential for spent fuel
pool criticality. : :

Appendix 3 A3-3


NEI

NUCLEAR ENERGY INSTITUTE

Lynnette Hendricks
DIRECTOR

PLANT SUPPORT

NUCLEAR GENERATION DIVISION

November 12, 1999

Richard J. Barrett

Chief, Probabilistic Safety Assessment Branch
U.S. Nuclear Regulatory Commission
Washington, DC 20555

Dear Mr. Barrett,

Industry is committed to performing decommissioning with the same high level of
commitment to safety for its workers and the public that was present during
operation of the plants. To that end, industry is making several commitments for
procedures and equipment which would reduce the probability of spent fuel poo!
events during decommissioning and would mitigate the consequences of those
events while fuel remains in the spent fuel pool. Most of these commitments are
already in place in the emergency plans, FSAR requirements, technical
specifications or regulatory guidance that decommissioning plants must follow.

These commitments were initially presented at the NRC public workshop on
decommissioning, July 15-16, in Gaithersburg, Maryland. They were further
discussed in detailed industry comments prepared by Erin Engineering. At a recent
public meeting with NRC management it was determined that a letter clearly
delineating these commitments could be useful to NRC as it considers input to its
technical analyses.

lam hereby transmitting those industry commitments as follows.

1. Cask drop analyses will be performed or single failure proof cranes will be
in use for handling of heavy loads (i.e., phase Il of NUREG 0612 will be
implemented).

2. Procedures and training of personnel will be in place to ensure that on site
and off-site resources can be brought to bear during an event. \c)o(

3. Procedures will be in place to establish communication between on site
and off-site organizations during severe weather and seismic events.

4. An off-site resource plan will be developed which will include access to
portable pumps and emergency power to supplement on site resources.
The plan would principally identify organizations or suppliers where off
site resources could be obtained in a timely manner.

5. Spent fue! pool instrumentation will include readouts and alarms in the


Response: Irradiation-induced degradation of steel requires high neutron fluency, which is not
present in the spent fuel pools. Over 30 years of operating experience has not indicated any
degradation of liner plates or the concrete that can be attributed to radiation effects.

With aging, concrete gains compressive strength of about 20% in an asymptotic manner and
the strength of reinforcing bars does not change with age, provided that rebars are not
degraded by corrosion. In general, degradation of concrete structures can be divided into two
parts, long term and short term. The long-term degradation can occur due to freezing and
thawing effects when concrete is exposed to outside air. This is the predominant long-term
failure mode of concrete, which is observed on bridge decks, pavements, and structures
exposed to weather. Degradation of concrete can also occur when chemical contaminants
attack concrete. These types of degradation have not been observed in spent fuel pools in any
of the operating reactors. Additionally, inspection and maintenance of spent fuel pool
structures are within the scope of the maintenance rule, 10 CFR 50.65, and corrective actions
are required if any degradation is observed. An inspection of the SFP structure to identify
cracks, spalling of concrete, etc., is also part of the seismic checklist. Substantial loss of
structural strength requires long-term corrosion of reinforcing steel bars and substantial
cracking of concrete. This is not likely to happen because of inspection and maintenance
requirements. Through the use of the seismic checklist, any degradation such as spalling of
concrete or cracks and indications of rust and stains, etc., will be detected and appropriate
corrective actions taken. (See Appendix 2b)

Degradation of the liner plate can occur due to cracks that can develop at the welded joints.
Seepage of water through minute cracks at welded seams has been minimal and has not been
observed at existing plants to cause structural degradation of concrete. Nevertheless,
preexisting cracks would require a surveillance program to ensure that structural degradation is
not progressing.

Based on the discussion above, any potential aging of the spent fuel pool structure is managed
during decommissioning. The structural strength will be verified using the seismic checklist in
the early stages of decommissioning, which may include site-specific analysis. While its
structural strength is not expected to degrade during decommissioning, it is managed under the
maintenance rule. As a result, the staff does not believe that detailed generic analysis is
needed.

Comment #53: To my knowledge, not every spent fuel pool was designed to the seismic
criteria in use today. The use of works like “robust” does not necessarily address seismic
qualifications. The NRC should identify all spent fuel pools that were not initially designed to
seismic criteria and explain their level of qualification, including the SF racks.

Response: When the licensee requests to expand the spent fuel storage capacity, spent fuel
pools undergo seismic and structural reevaluation during a licensing review. Spent fuel pool
structures, as well as the spent fuel racks, undergo detailed analysis and staff review. All
currently operating nuclear power plants have expanded their spent fuel storage capacity and
met their safe shutdown earthquake criteria.

Comment #54: Not all PWR buildings housing spent fuel are seismically qualified. The NRC
should perform a worst case analysis of the result of a seismic event which collapses the spent
fuel pool building, and/or drains the pool and/or damages the spent fuel. Both criticality and
zirconium fires are of concern. The nine initiating events listed on p. 11 which could occur

Appendix 6 AG-16


Operating reactors are not required to implement Phase Il of NUREG-0612. The risk for SFPs
at operating plants is limited by a lower expected frequency of heavy load lifts then at
decommissioning plants. Nonetheless, this issue will be further examined as part of the Office
of Nuclear Regulatory Research’s prioritization of Generic Safety Issue 186, “Potential Risk and
Consequences of Heavy Load Drops in Nuclear Power Plants,” which was accepted in May
1999.

3.4.7 Spent Fuel Pool Uncovery Frequency at Times Other Than 1 Year After Shutdown

The staff has considered how changes in recovery time available to fuel handlers at 2 months,
2 years, 5 years, and 10 years after shutdown (see Table 2.1) change the insights or bottom-line
numerical results from the risk assessment. The different recovery times primarily affect the
human reliability analysis (HRA) results and insights. Even at 2 months after shutdown, the
HRA failure estimates are small and are dominated by institutional factors (e.g., training, quality
of procedures, staffing). It is therefore expected that the total fuel uncovery frequency at

2 months will continue to be dominated by the seismic contribution. At periods beyond 1 year,
the increased recovery time (from a very long time to an even longer time) lowers the
uncertainty that these HRA estimates really are very small, but the increased time has not
translated into significant changes in the bottom-line numerical estimates because quantification
of the effect of such extensions on organizational problems is beyond the state-of-the-art.

3.5 Beyond Design Basis Spent Fuel Pool Accident Scenarios (External Events)

In the following sections, the staff explains how each of the external event initiators was
modeled, discusses the frequency of fuel uncovery associated with the initiator, and describes
the most important insights regarding risk reduction strategies for each initiator.

3.5.1 Seismic Events

The staff performed a simplified seismic risk analysis in its June 1999 preliminary draft risk
assessment to gain initial insights on seismic contribution to SFP risk. The analysis indicated
that seismic events could not be dismissed on the basis of a simplified bounding approach. The
additional efforts by the staff to evaluate the seismic risk to SFPs are addressed here and in
Appendix 2B.

SFP structures at nuclear power plants should be seismically robust. They are constructed of
thick, reinforced concrete walls and slabs lined with stainless steel liners 1/8 to 1/4 inch thick.?
Pool walls are about 5 feet thick and the pool floor slabs are around 4 feet thick. The overall
pool dimensions are typically about 50 feet long by 40 feet wide and 55 to 60 feet high. In
boiling-water reactor (BWR) plants, the pool structures are located in the reactor building at an
elevation several stories above the ground. In pressurized-water reactor (PWR) plants, the SFP
structures are outside the containment structure and supported on the ground or partially

Except at Dresden Unit 1 and Indian Point Unit 1, have no liner plates. The plants were
permanently shut down more than 20 years ago and no safety significant degradation of the
concrete pool structure has been reported.

3-18


For Robinson (PWR), Ref. 1 states that the critical failure mode is an out-of-plane
bending failure of the East wall. With this failure mode, the liner will be breached and
the concrete will become rubbled over a zone equal to the wall thickness at the base of
the wall and along the two sides (ends) of the wall. The outward flow of water is likely to
be somewhat slower than for a shear crack, but is still expected to be rapid. Probably less
water will be retained in the pool than for the case of a shear crack through the wall, and
more water will be retained than for the case of a shear crack through the pool floor.

Although not reported as the critical failure mode in Ref. 1, I believe that either of
the two shear failure modes reported above for a BWR could also be the critical failure
mode for some PWR pools.

Lastly, for stronger spent fuel pools with greater out-of-plane flexure and shear
capacities, an in-plane shear failure mode for one or more of the pool walls could control.
I suspect this will be the case for particularly some PWR pools. With this failure mode,
the liner will be breached and the concrete wall will be cracked in a diagonal X pattern of
cracking from near the base of the wall at the edges to near the top of the wall at the
opposite edges. The pool will empty to near the base of the wall with probably some
small amount of water being retained in the pool.

No matter which of these failure modes occur, drainage of the pool is expected to
be fairly rapid. A small, but uncertain, amount of water is likely to remain in the pool
with post-seismic-failure water depths ranging from essentially zero depth to about 4-feet
of depth depending upon the critical failure mode.

3. Expected Level of Collateral Damage

The seismic capacity of spent fuel pools is high. For spent fuel pools that have
successfully passed the NEI/NRC seismic walkdown procedure, I believe the spent fuel
pool will have at least about the following seismic fragility capacities:

Spent Fuel Pool
Cy, = 05g PGA
Cio, = 0.75g PGA qd)
Coo, = 1.25g PGA

where Cys, Cig, and Csoy, are the 1%, 10%, and 50% non-exceedance probability (NEP)
peak ground acceleration capacities.

For the Central and Eastern U.S. (CEUS), I estimate the following seismic
fragilities:

A2B-24


consequence analysis indicated that zirconium cladding fires could give offsite doses that the
NRC would consider unacceptable. To assess the risk during the period of vulnerability to
zirconium cladding fires, the staff initially performed a broad preliminary risk assessment, which
modeled many internal and external initiating events. The preliminary risk assessment was
made publicly available early in the process (June 1999 [Ref. 3]) so that the public and the
nuclear industry could track the NRC’s evaluation and provide comments. In addition, the
preliminary risk assessment was subjected to a technical review and requantification by the
Idaho National Engineering and Environmental Laboratory (INEEL). The NRC continued to
refine its estimates, putting particular emphasis on improving the human reliability assessment
(HRA), which is central to the analysis given the long periods required for lowering the water in
the SFP for most initiators. The staff identified those characteristics that a decommissioning
‘plant and its utility should have to assure that the risks driven by fuel handler error and
institutional mistakes are maintained at an acceptable level. In conjunction with the staffs HRA
effort and ongoing reassessment of risk, the nuclear industry through NE! developed a list of
commitments (NEI letter dated November 12, 1999 [Ref. 4]) that provide boundaries within
which the risk assessment’s assumptions have been refined. The staff released a draft risk
assessment in February 2000, which updated the June 1999 preliminary study. The
assessment reflects the commitments made by industry, the additional requirements we have
developed to ensure the assumptions in the assessment remain valid, the technical review by
INEEL, the staff's ongoing efforts to improve the assessment, and input from stakeholders. The
study provides a technical basis for determining the acceptability of exemption requests and
future rulemaking on decommissioning plant risk.

The staff looked at the broad aspects of the issue. A wide range of initiators (internal and
external events including loss of inventory events, fires, seismic, aircraft, and tornadoes) was
considered. The staff modeled a decommissioning plant's SFP cooling system based on the
sled-mounted systems that are used at many current decommissioning plants. One
representative SFP configuration (see Appendix 2A, Figure 2.1) was chosen for the evaluation
except for seismic events, where the PWR and BWR SFP designs (i.e., the difference in location
of the pools in PWRs and BWRs) were specifically considered. Information about existing
decommissioning plants was gathered from decommissioning plant project managers and
during visits to four sites covering all four major nuclear steam supply system vendors (General
Electric, Westinghouse, Babcock & Wilcox, and Combustion Engineering). Plant visits gathered
information on the as-operated, as-modified SFPs, their cooling systems, and other support
systems.

From the perspective of offsite consequences, the staff focused on the zirconium fire end state,
because there has to be an energetic source (e.g., a large high temperature fire) to transport the
fission products offsite in order to have potentially significant offsite consequences. The staff
chose the timing of when the SFP inventory is drained to within three feet of the top of the spent
fuel as a surrogate for onset of the zirconium fire because once the fuel is uncovered, the dose
rates at the edge of the pool would be in the tens of thousands of rem per hour, because it is
unclear whether hydrides could cause ignition at lower cladding temperatures than previously
predicted, because of the differences in configurations, and because there was great difficulty in
modeling the heat transfer rate as the fuel was uncovered. In addition, from the point of view of
estimation of human error rates, since for initiating events (other than seismic and heavy load
drop) would take many days to uncover the top of the fuel, it was considered of small numerical

Appendix 2 A2-2



failure of a support system leading to fuel uncovery and a zirconium fire, a tornado is bounded
by other more probable events. Recovery of the support system(s) will reduce the likelihood of
spent fuel uncovery.

Missiles generated by high winds (for example, straight winds or hurricanes) are not as powerful
as those generated by tornados. Therefore high winds are estimated to have a negligible
impact on the frequency of catastrophic failure of the SFP resulting in fuel uncovery. Long-term
loss of offsite power due to straight winds is evaluated in Section 3.4.4.

The staff estimated the frequency of significant damage to SFP support systems from straight-
line winds to be very low. Damage was assumed to be caused by building collapse. Based on
the construction requirements for secondary containments, the staff believes that the buildings
containing BWR spent fuel pools are sufficiently robust that straight line winds will not challenge
the integrity of the building. The staff assumes buildings covering PWR spent fuel pools have a
concrete foundation that extends part way up the side of the building. The exterior of the rest of
the building has a steel frame covered by corrugated steel siding. The PWR spent fuel buildings
are assumed to be constructed to American National Standards Institute (ANSI) or American
Society of Civil Engineers (ASCE) standards. Based on these assumptions, the staff believes
that straight-line winds will cause buildings housing PWR spent fuel pools to fail at a frequency
of 1x10* per year or less. This failure rate for support systems is subsumed in the initiating
event frequency for loss of offsite power from severe weather events. The event tree for this
initiator takes into account the time available for recovery of spent fuel pool cooling
(approximately 195 hours for 1-year old PWR fuel and 253 hours for 1-year-old BWR fuel).

3.6 Criticality in Spent Fuel Pool

In Appendix 3, the staff performed an evaluation of the potential scenarios that could lead to
criticality and identified those that are credible.

In this section the staff gives its qualitative assessment of risk due to criticality in the SFP,
concluding that, with the additional assumptions, the potential risk from SFP criticality is small.

Appendix 3 references the NRC staff report “Assessment of the Potential for Criticality in
Decommissioned Spent Fuel Pools.” The assessment identified two credible scenarios listed
below:

(1) A compression or buckling of the stored assemblies from the impact of a dropped heavy
load (such as a fuel cask) could result in a more optimum geometry (closer spacing) and
thus create the potential for criticality. Compression is not a problem for high-density PWR
or BWR racks because they have sufficient fixed neutron absorber plates to mitigate any
reactivity increase, nor is it a problem for low-density PWR racks if soluble boron is
credited. But compression of a low-density BWR rack could lead to a criticality since BWR
racks contain no soluble or solid neutron-absorbing material. This is not a surprising result
since low-density BWR fuel racks use geometry and fuel spacing as the primary means of
maintaining subcriticality. High-density racks rely on both fixed neutron absorbers and
geometry to control reactivity. Low-density racks rely solely upon geometry for reactivity
control. In addition, all PWR pools are borated, whereas BWR pools contain no soluble

3-25


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to be met for as long as nuclear materials remain on the plant site. Accordingly, the staff has
compared the risks to an individual with the QHOs, assuming the licensee maintains the facility
at the recommended PPG of 1x10° per year. The relevant risk measures are the early fatality
risk to an average individual within 1 mile of the plant, and the latent cancer fatality risk to an
average individual within 10 miles of the plant. These measures would not be significantly
impacted by population density since they are determined on the basis of the risk to the
average individual.

Appendix 4B of this study provides the results of offsite consequence calculations for an SFP
fire occurring at various times following final shutdown at a hypothetical 3441 MWth BWR SFP
located at the Surry site. Additional calculations provided in Appendix 4A address the
sensitivity of early and latent health effects to source terms, time of evacuation, percent of
population participating in the evacuation, population distribution, number of cores participating
in the SFP fire, and plume-related parameters. The risk measures corresponding to the above
numerical objectives were calculated using the MACCS2 computer code for each of the cases
Teported in Appendix 4B (i.e., low and high ruthenium source term with both early and late
evacuation), and for the worst case SFP accident source term reported in Appendix 4A. The
latter case, identified as Case 45a, corresponds to a complete release of the volatiles (cesium
and iodine) and ruthenium, a 0.01 release of fuel fines, and late evacuation of 95 percent of the
population. The results are reported in Table 1. For comparison with the numerical objectives,
the staff assumed that the licensee maintains the facility at the recommended PPG of 1x10°
per year.

The risk results indicate that at a PPG of 1x10° per year, the QHOs would continue to be met
for even the most severe cases considered in Appendix 4A and 4B. The margins to both
QHOs are substantial (about two orders of magnitude) for the case with early evacuation even
with the large ruthenium release. The margins are considerably reduced in the late evacuation
cases, but sufficient to conclude that the QHOs would be met given the bounding nature of the
source terms and fission product inventories used in these calculations. Although the
comparisons in Table 1 are at a time 1 year after shutdown, the staff also evaluated the risk at
30 days after shutdown and found that it continues to meet the QHOs.

The margin to the QHO is smallest (i.e., the percent of QHO is the largest) for early fatality risk.
Thus, similar to severe accidents in operating reactors, acceptable levels of risk for an SFP
accident would be controlled by the early fatality risk measure. The margins to the QHO
observed in these calculations suggest that the recommended PPG of 1x10° per year provides
an appropriate level of safety.

The role of the PPG in plant-specific implementation of regulatory changes for
decommissioning plants will be established as part of the integrated rulemaking. In one
possible approach shown in Figure 1, a licensee that fully complies with all IDCs and SDAs
(including the seismic checklist) might be permitted to implement changes under the revised
rule without a plant-specific analysis and detailed staff review. However, if the licensee/site
does not comply with all of the IDCs and SDAs, a plant-specific analysis of SFP risks would be
required in order to support relaxations to existing regulations. The PPG could be used to
establish an acceptable level of risk in the review of such submittals.

2. CONCLUSIONS

The frequency of events leading to uncovery of the spent fuel must be less than 1x10° per year
in order to consider risk-informed changes that could result in the equivalent of a 1x10° per
year increase in LERF. Based upon the above comparisons, the staff believes that the LERF-
based poo! performance criteria of 1x10° per year is reasonable and appropriate. This is
supported by the comparisons that show that the conditional health effects for SFP fires may be
in the range of health effects considered for severe accidents in operating reactors, and that

Appendix 4C A4C-3


5. SUPPORT SYSTEM AVAILABILITY

An F2 or larger tornado would be needed to consider damage to support systems ( power
supplies, cooling pumps, heat exchangers, and makeup water sources). The likelihood of the
exceedance of this size tornado is estimated to be 1.5x10° per year (for Region A), or lower.
This frequency is bounded by other more likely initiators that can cause loss of support
systems.

6. REFERENCES

4 NUREG/CR-2944, "Tornado Damage Risk Assessment,” Brookhaven National Laboratory,
September 1982

2 http:/Awww.ncde.noaa.gov/

3 NUREG/CR-5042, "Evaluation of External Hazards to Nuclear Power Plants in the United
States," Lawrence Livermore National Laboratory, December 1987.

4 DOE-STD-1020-94, "Natural Phenomena Hazards Design and Evaluation Criteria for
Department of Energy Facilities," January 1996, Department of Energy

Appendix 2E A2E-4



Hardware Failure Probabilities

Unavailability of an SFP makeup system, SFP-REGMKUP-F, was assigned a value of 5.0E-2,
using the estimate from INEL-96/0334. It is assumed that the SFP makeup system is
maintained since it is required often to provide make-up.

4.5.6.4 Basic Event Proababilities

Basic Event Basic Event Probability
HEP-MKUP-START-E 2.5E-4
HEP-MKUP-START 2.5E-6
SFP-REGMKUP-F 5.0E-2

4.5.7 Top Event OMK - Operator Initiates Make-up Using Fire Pumps

4.5.7.1 Event Description and Timing

This top event represents failure to provide make-up using the firewater pumps. The case of a

large leak is represented by a fault tree LOI-OMK-LGLK. In this case the operators have 40

hours to start a firewater system. The case of a small leak is represented by two functional fault

trees, LOI-OMK-SMLK, and LOI-OMK-SMLK-L. The difference between the two trees is that in
the first, the operators are aware of the problem and are attempting to solve it, whereas in the
second, the operators will need to first recognize the problem. In both small leak cases, the
operator has more than 65 hrs to start a firewater system. In all cases neither of the firewater
pumps would be initially unavailable.

4.5.7.2 Relevant Assumptions,

. The operators have 40 to 65 hours to start a firewater pump depending on the leak size.

. There is a means to remotely align a makeup source to the SFP without entry to the refuel
floor so that make-up can be provided even when the environment is uninhabitable
because of steam and/or high radiation (NEI commitment no.8).

. Repair crew is different than onsite operators.

. On average, it takes 10 hours to repair a pump if it fails to start and run.

. It takes 16 hours to contact maintenance personnel, make a diagnosis, and get new parts.

. Both firewater pumps are located in a separate structure and are protected from the
potential harsh environment in the case of pool bulk boiling.

. Maintenance and testing are performed on diesel-driven and electric firewater pumps to
maintain operable status (NEI commitment no. 10).

Appendix 2A A2A-60


hypothetical accidents that are often termed "incredible." Also, the consequences of such
accidents need to be considered. The staff has previously stated that while it is correct that the
frequency of events that could lead to a zirconium fire is small, the consequences of such a fire
could be significant. The SRM for SECY-93-127 approved the staff's recommendation that
after an appropriate cooling period for the spent fuel had elapsed that primary level coverage
could be reduced and licensees would be allowed to withdraw from participation in the
secondary financial protection layer. The actual amount of coverage and acceptable methods
to demonstrate approval will be determined during rulemaking.

Comment #81: With new personnel and decommissioning personnel, what methods are
available to instill or ensure the same “safety culture” as during operation?

Response: There are several methods of instilling/ensuring “safety culture” in new personnel at
both operating and decommissioning facilities. Methods include management policies and
procedures, training, and qualification. OSHA requires employers to provide employees with
safety training and education. Section 1926.21(b)(2) of Title 29 of the CFR requires training in
the recognition and avoidance of unsafe conditions, 29 CFR 1926.21(b)(3) requires training in
the safe handling and use of poisons, caustics, and other harmful substances, 29 CFR
1926.21(b)(5) requires training in the safe handling and use of flammable liquids, gases, or
toxic materials, and 29 CFR 1926.21(b)(6) requires confined or enclosed space training. In
addition, 10 CFR 50.120 requires training and qualification of nine categories of personnel
involved with spent fuel pool maintenance and support. The training programs for the nine
categories of personnel should include occupational safety and radiation protection training.
While NRC and OSHA require training, it is incumbent upon the licensee to provide the training
and instill/ensure upon the workers the proper “safety culture.”

Comment #82: A commenter asked about calculations for radiation dose experienced by
members of the fire brigade responding to resin fires.

Response: Existing regulatory requirements address the need for onsite worker radiation
protection and emergency plans to consider protective actions and a means for controlling
exposures in an emergency for emergency workers as well as the public. For example, the
regulatory requirements for emergency worker protective actions and exposure control are
found in 10 CFR 50.47(b)(10) and 10 CFR 50.47(b)(11). Each site has established procedures
and training for the protection of workers responding to emergency situations. Generally, these
procedures include the consideration of radiological conditions when responding to events.
Calculations for occupational exposure for emergency workers would be consistent with the
EPA Emergency Worker and Lifesaving Activity Protective Action Guidelines.

Appendix 6 A6-27


miles. The MACCS results in terms of these two consequence measures are given in Table 10.

Table 10 Results of Release Fraction Sensitivities
(95% evacuation, Surry Population Density)

Case | Release Fraction Mean Consequences
ICs | Ru Te | Ba Sr Ce La Individual Risk Individual Risk

of an Early of a Cancer
Fatality (within Fatality (within
‘one mile) ten miles)

45a 1 1 02 | .01 .01 01 .01 3.66x107 5.16x10

45b 75 .02_ | .01 -01 01 .01 3.23x107 4.98x107

46a? 41 1 02 | .01 01 01 01 41.61x10* 2.83x103

46b? 78 75 02 | .01 01 01 .01 1.40x10* 2.55x10*

*Based on evacuation before release.
Amount of Fue! Releasing Fission Products:

To assess the sensitivity to the fission product inventory released, the staff performed calculations with
all of the spent fuel (i.e., 3.5 cores) and the final core offload releasing fission products. These
calculations were run for cases with evacuation beginning after the release begins. The inventories used
in the MACCS caiculations for one core are the Table A.5 inventories in the BNL study reduced by one
year of radioactive decay. The results of the MACCS calculations are given in Table 11.

Table 11 Sensitivities on Amount of Fuel Assemblies Releasing Fission Products

(99.5% evacuation)
Case Population Ruthenium | #of Mean Consequences (within 100 miles)
Density Release cores
Fraction Prompt Societal Cancer
Fatalities Dose Fatalities

(person-

rem)
Base Case Surry 2x10° 3.5 1.01 4.54x10® 2,320
34 Surt 2x10 4 014 3.23x10® 1,530
1 Surry 1 3.5 95.3 9.53x10° 9,150
32 Surry 4 1 50.5 7.25x10° 7,360
21 uniform 2x10 3.5 9.33 5.05x10® 2,490
33 uniform 2x10° 1 77 3.10x10® 1,480
22 uniform 1 3.5 134 9.46x10° 6,490
34 uniform 4 1 103 6.59x10® 4,960

For the cases with a ruthenium release fraction of 2x10°, the reduction in prompt fatalities is caused by
the reduction in the Cs-137 inventory which decreases from 8.38x10"” Bq to 2.11x10" Bq in going from
3.5 cores to one core. This was confirmed by rerunning Case 33 with a Cs-137 inventory of 8.38x10”

-8-


4.0 IMPLICATIONS OF SPENT FUEL POOL (SFP) RISK FOR REGULATORY
REQUIREMENTS

The primary purpose of this study is to provide risk insights to support possible revisions to
regulatory requirements for decommissioning plants. Section 4.1 below describes the safety
principles of Regulatory Guide (RG) 1.174 as they apply to an SFP, and examines the design,
operational, and regulatory elements that are important in ensuring that the risk from an SFP
continues to meet these principles. This technical assessment explores possible implications for
EP requirements, but the same technical information also provides risk insights to inform
regulatory decisions on changes in insurance, safeguards, staffing and training, backfit, and
other requirements for decommissioning plants. Section 4.2 examines the implications of the
technical results for these regulatory decisions, and how future regulatory activity might reflect
commitments and assumptions. The implications of safeguards events are not included in this
evaluation.

4.1 Risk-Informed Decision Making

In 1995, the NRC published its PRA policy statement (Ref. 1), which stated that the use of PRA
technology should be increased in all regulatory matters to the extent supported by the state-of-
the-art of the methods. Subsequent to issuance of the PRA policy statement, the agency
published RG 1.174, which contained general guidance for application of PRA insights to

the regulation of nuclear reactors. The regulatory framework proposed in this study for
decommissioning plants is based on the risk-informed decision-making process described in

RG 1.174 (Ref. 2). Although the focus of RG 1.174 is decision making regarding changes to the
licensing basis of an operating plant, the same risk-informed philosophy can be applied
generically as part of the evaluation of potential exemptions or changes to current regulatory
requirements for decommissioning plants.

RG 1.174 articulates the following safety principles, which can be applied in evaluating
regulatory changes for decommissioning plants:

* The proposed change meets the current regulations unless it is explicitly related to a
requested exemption or rule change, i.e., a “specific exemption” under 10 CFR 50.12 ora
“petition for rulemaking” under 10 CFR 2.802.

* When proposed changes result in an increase in core damage frequency and/or risk, the
increases should be small and consistent with the intent of the Commission's safety goal
policy statement.

+ The proposed change is consistent with the defense-in-depth philosophy.

+ The proposed change maintains sufficient safety margins.

+ The impact of the proposed change should be monitored using performance measurement
strategies.

4-1


ttl='w15x15 in High Density Rack’
xul=-11.5 yul= 11.5 zul=0.0

xlr= 11.5 ylr=-11.5 zlr=0.0

uax=1 vdn=-1 nax=750

end plot
end data
end
Table 1 Eigenvalue (using infinite multiplication factor) reduction from skipping cells
between high reactivity assemblies.
Skipped Cells PWR BWR
2 1.03533 1.02628
4 : 1.01192 * 4.01503
6 1.00363 1.01218
8 0.99786 1.01059

Appendix 3 A3-13


analysis would be required for the Western sites and Robinson for the following reasons: (1) the
checklist helps assure an SFP has adequate capacity at the 1.2g PSA ground motion level,

(2) at these sites the frequency of ground motions that exceed 1.2g PSA may be so high as to
call into question whether the risk at a decommissioning plant would exceed the NRC’s Safety
Goals, and (3) plant-specific seismic analyses may demonstrate that the sites actually have
adequate margin to meet the staffs PPG and Safety Goals.

The staff determined the mean of all the LLNL hazard estimates for operating reactor sites east
of the Rocky Mountains (i.e., 2x10* per year). This value is a factor of five less than PPG and
the mean bounds about 70 percent of the sites east of the Rockies. Similarly, the mean of all
EPRI hazard estimates is 2x10” per year. From a risk standpoint, the majority of potential
decommissioning sites have expected frequencies of fuel uncovery far below the staff's pool
performance guideline. Risk results are reported in Section 3.7 of the main study.

3. SEISMIC CHECKLIST

The staff determined that, absent specific information about SFP seismic capacities, some plant-
specific evaluation of SFP capacity was warranted. During stakeholder interactions with the
staff, the staff proposed the use of a seismic checklist that built on the work done to quantify
seismic margins and that could provide assurance of the capacity of SFPs. In a letter dated
August 18, 1999, NEI proposed a checklist that could be used to show robustness for a seismic
ground motion with a peak spectral acceleration (PSA) of 1.2 g (or with peak ground
acceleration (PGA), which is not as good an estimator, of approximately 0.5 g). This checklist
was reviewed and enhanced by the staff (see Attachment 1). Dr. Robert Kennedy, a staff
consultant, reviewed the enhanced checklist and concluded that the screening criteria are
adequate for the vast majority of central and eastern U.S. sites. The seismic checklist was
developed to provide a simplified method for demonstrating a HCLPF at an acceptably low
value of seismic risk. The checklist includes elements to assure there are no weaknesses in the
design or construction nor any service induced degradation of the pools that would make them
vulnerable to failure under earthquake ground motions that exceed their design basis ground
motion but are less than the HCLPF value. SFPs that satisfy the seismic checklist, as written,
would have a high confidence in a low probability of failure for seismic ground motions up to

1.2 g peak spectral acceleration.

4. SEISMIC RISK - SUPPORT SYSTEM FAILURE

In its preliminary draft study published in June 1999, the staff assumed that a ground motion
three times the SSE was the HCLPF of the SFP. The HCLPF capacity is such that 95 percent
of the time the pool would remain intact (i.e., would not leak significantly given ground motion up
to a certain value, i.e., 1.2g PSA). The staff evaluated what would happen to SFP support
systems (i.e., the pool cooling and inventory make-up systems) in the event of an earthquake
three times the SSE. The staff modeled some recovery as possible (although there would be
considerable damage to the area’s infrastructure at such earthquake accelerations). The
estimate in the preliminary study for the contribution from this scenario was 1x10° per year. In
this study, this estimate has been refined based on looking at a broader range of seismic
accelerations and further evaluation of the conditional probability of recovery under such
circumstances. The staff estimates that for an average site in the northeast U.S. the return

Appendix 2B A2B-3


Plume-Spreading Point in Early Fatalities Societal Dose Cancer Fatalities

Model Distribution (rem)

default not applicable : 95.3 9.53x10® 9,150

NUREG/CR-6244 10" percentile 527 9.04x10® 8,343
50" percentile 8.89 1.26x10" 10,100
mean 54.4 1.28x10" 10,100
90" percentile 171 4.66x107 11,900

Table 14 Results of Plume-Spreading Model Sensitivity for Case 14
(99.5% evacuation, Surry Population Density)

Plume-Spreading Point in Early Fatalities Societal Dose Cancer Fatalities

Model Distribution (rem)

default not applicable 132 6.75x10® 6,300

NUREG/CR-6244 10" percentile .00197 7.00x10® 6,010
50" percentile .00855 1.03x107 7,730
mean 118 1.07x107 7,810
90" percentile 0637 1.46x107 9,590

Conclusion

Appendix 4 documents the staffs evaluation of the offsite consequences of a spent fuel pool accident
involving a sustained loss of coolant, leading to a significant fuel heatup and resultant release of fission
products to the environment. The objectives of the staffs evaluation were (1) to assess the effect of one
year of decay and (2) to assess the effect of early versus late evacuation because spent fuel pool
accidents are slowly evolving accidents. The staff's evaluation was an extension of an earlier study
performed by BNL for spent fuel pools at operating reactors, which assessed consequences using
inventories for 30 days after shutdown. Subsequent reviews of the staffs consequence evaluation
identified issues with the earlier evaluation performed by BNL in the areas of fission product source term
and plume-related parameters. To address these issues, the staff performed additional MACCS.
sensitivity calculations which are documented in the current appendix.

With regard to the fission product source term, sensitivity calculations were performed using different
release fractions for the nine fission product groups. These calculations also included variations in
Population density, evacuation start time, percentage of the population evacuating, and number of fuel
assemblies releasing fission products. With regard to plume-related parameters, sensitivity calculations
were performed using different plume heat contents and updated values for the plume-spreading
parameters. .

With the exception of ruthenium, increasing the release fraction of each fission product group resulted in
a negligible to modest (less than 40%) increase in consequences. Increasing the ruthenium release
fraction resulted in a larger increase in consequences. However, these consequence increases were
demonstrated to be largely offset by beginning the evacuation before the release begins. Such an early
evacuation is likely, because after a year of decay, it will take a number of hours for the fuel with the
highest decay power to heat up to the point of releasing fission products.

Other sensitivity calculations involved examining the effect of (1) decreasing the amount of fuel releasing
fission products from the entire spent fuel pool inventory to the final core offload and (2) decreasing the

-11-


4.4.2.2 Relevant Assumptions
. See section 4.4.2.3 below.
4.4.2.3 Quantification

Non-HEP Probability

NUREG-1032 (Ref. 19) classified LOSP events into plant-centered, grid-related, and
severe-weather-related categories, because these categories involved different mechanisms
and also seemed to have different recovery times. Similarly, NUREG/CE-5496 divides LOSP
events into three categories and estimates different values of non-recovery as functions of time.
A non-recovery probability within 24 hrs for the offsite power from the severe weather event was
estimated to be 2.0E-2 to <1.0E-4 depending on the location of the plant. In the operating plant,
recovery of offsite power may be very efficient because of presence of skilled electricians. In
the decommissioned plant, the skilled electricians may not be present at the site. Therefore, for
the purpose of this analysis, a non-recovery probability for offsite power because of severe
weather event (REC-OSP-SW) of 2.0E-2 is used.

4.4.2.4 Basic Event Probability

Basic Event Basic Event Probability
REC-OSP-SW 2.0E-2

4.4.3. Top Event OCS — Cooling System Restart and Run
4.4.3.1 Event Description and Timing

This top event represents restarting the SFP cooling system, given that offsite power has been
recovered within 24 hours. There are two electrically operated pumps and the operator can
start either one. If the operator starts the pump that was in operation, no valve alignment would
be required. However, if operator starts the standby pump, some valve alignment may be
required.

Fault tree LP2-OCS has several basic events: an event representing failure of the operators to
realize they need to start the SFP cooling system, an operator action representing the failure to
establish SFP cooling, and several hardware failures of the system. If power is recovered within
24 hours, the operator has 9 hours to start the system before boil-off starts. If he fails to initiate
SFP cooling before boil-off begins, the operator must start a firewater pump to provide make-up.

Appendix 2A A2A-44


ze eunbi4

uo}}e19]999¥y [21}9Edg 4eOd Z'| JO Jd TOH Pu SO}EUISA NTT
uo peseg ane, peonpul Ayjeowisias joog jan4 yueds jo Aouenbel4

80-51

y L0-Sb

+ 90-21

GO-AL

~ poral

(aeak 49d) Aouenbesy eanjiey

3-21


multiple cores may be involved; the relevant cladding and fuel degradation mechanisms could
lead to increased releases of certain isotopes (e.g., short-lived isotopes such as iodine will have
decayed, but the release of long-lived isotopes such as ruthenium could be increased due to air-
fuel reactions); and there is no containment surrounding the SFP to mitigate the consequences.
On the other hand, after about 2 years, the consequences are different than from a large early
release because the postulated accidents progress more slowly, allowing time for protective
actions to be taken to significantly reduce early fatalities (and to a lesser extent latent fatalities).
In effect, an SFP fire would result in a “large” release, but this release would not generally be
considered “early” due to the significant time delay before fission products are released.

In spite of the differences relative to an operating reactor large early release event and the
differences in isotopic makeup, the consequence calculations performed by the staff and
discussed in Section 3.7 show that SFP fires could have health effects comparable to those of a
severe reactor accident. These calculations considered the effects of different source terms and
evacuation assumptions on offsite consequences. Since an SFP fire scenario would involve a
direct release to the environment with significant consequences, the staff has decided that the
RG 1.174 guidance concerning LERF can be applied to the issue of SFP risks for
decommissioning plants.

The LERF guidance is applied in two ways in this study:

(1) Because the changes in EP requirements affect not the frequency of events involving a
large early release (i.e., the SFP fire frequency) but the consequences of these releases,
the allowable increase in LERF in RG 1.174 is translated into an allowable increase in key
risk measures. The estimated risk increases associated with changes in EP requirements
are then compared to the allowable increases inferred from RG 1.174. These comparisons
are presented in Appendix 4D.

(2) The RG 1.174 guidance is used to establish a PPG. The PPG provides a threshold for
controlling the risk from a decommissioning SFP. By maintaining the frequency of events
leading to uncovery of the spent fuel at a value less than the recommended PPG value of
1x10° per year, zirconium fires will remain highly unlikely, the risk will continue to meet the
Commission’s QHOs, and changes to the plant SFP licensing basis that result in very small
increase in risk may be permitted consistent with the logic in RG 1.174. A licensee would
need to assure that the frequency of events leading to uncovery of the spent fuel would be
less than the PPG in order to consider the risk-informed changes in a rule for
decommissioning plants. With the exception of those plants mentioned above, this
assurance could be provided by conforming with the IDCs and SDAs listed in Tables 4.2-1
and -2. The use of the LERF guidance (1x10° per year frequency of fuel uncovery) was
questioned by the ACRS because of concerns related to SFP source terms and accident
consequences. The rationale for the PPG is presented in Appendix 4C.

The risk increases associated with relaxations in EP requirements compare favorably with the
guidance contained in RG 1.174 (see Table 4 of Appendix 4D). Relaxation of EP requirements
would result in an increase of about 1.5x10° early fatalities per year and 2 person-rem per year
for the Surry analysis, the first is about a factor of 15 and the second a factor of 5 below the
allowable increase inferred from the RG 1.174 LERF criteria. The increase in the QHO risk

44


FAILURETOISOLATE
LOSS AND INTIATE
(EARLY RESPOND)
by ISU
— t 1
OPERATOR FAILS REGULAR SFP OPERATOR FALS TO
TOISCLATE LEAK MAKEUP SYSTEM START MAKUP_
FAILS: (EARLY RESPOND)
HEP-LEAKISO. oO. HEP-MKUR.-START-£
i
FAILURE TO INITIATE
NORMAL MAKEUP
(SMALL LEAK)
Lo} ‘ LU
I 1
REGULAR OPERATOR FAILS TO
MAKEUP ‘SSrem ‘START MAKUP
FAILS (SMALL LEAK)
SFP-REGMKUP-F HEP-MKUP-START

A-17



Table 9 Results of Release Fraction Sensitivities
(95% evacuation, Surry Population Density)

Case | Release Fraction Mean Consequences (within
100 miles)
ICs | Ru Te |Ba Sr Ce La Early Societal | Cancer
Fatali- | Dose Fatalities
ties (person-
rem)
1 1 2x10% | .02 | .002 | .002 | 1x10 | 1x10* 1.04 4.54x1Q 2,320
45 1 1 02 | .002 | .002 | 1x10* | 1x10* 92.2 9.50x1Q 9,150
45a 1 41 02 | .01 01 .01 01 103 1.33x19 11,700
45b 75 75 02 | .01 01 .01 01 54.9 1.17x19 10,300
46° 4 1 02 | 002 | .002 | 1x10* | 1x10* 1.32 | 6.84x1Q 6,430
46a? 1 1 02 | .01 01 01 01 1.54 8.89x1Q 8,160
46b* 75 is) 02 | .04 01 01 01 543 | 7.94x1Q 6,880 ©
46c? 75 75 75 | 01 01 .01 01 544 7.94x1Q 6,880
46d? 75 75 75 | 75 01 -01 01 544 7.94x1Q 6,880
46e° 75 75 75 | 75 75 01 01 644 1.01x19 8,350

“Based on evacuation before release.

In addition, the staff performed calculations with 95% of the population evacuating with the evacuation
beginning three hours before the release begins. The results of these calculations are shown in Table 9.
The starting point for these calculations was Case 46, which includes evacuation beginning three hours
before the release begins. Then, Case 46a was run with a fuel fines release fraction of .01. The same
trends were seen as in the 99.5% evacuation cases, Cases 14 and 97.

The main difference between the results for 99.5% and 95% evacuation is in the area of early fatalities
for cases with evacuation before release. In comparing Cases 14 and 97 with Cases 46 and 46a, a
factor-of-ten increase in early fatalities is seen, because of the factor-of-ten increase in persons not
evacuating. Cases 14 and 97 use one out of 200 people not evacuating, while Cases 46 and 46a use
ten out of 200 people not evacuating.

The staff also performed sensitivity calculations for tellurium, barium, and strontium by increasing their
release fractions to that of the volatile fission products, that is, .75. In Case 46c, the release fraction for
tellurium was increased from .02 to .75. In Case 46d, the release fraction for barium was increased from
.01 to .75. No change in consequences were seen in these two cases, because of the small inventories
of these isotopes after a year of decay. In Case 46e, the release fraction for strontium was increased
from .01 to .75. A-small increase in the consequences was seen in this case.

The results in Table 9 are the total number of early fatalities, societal dose, and cancer fatalities for the

population within 100 miles of the facility. However, the NRC’s quantitative health objectives are given in
terms of individual risk of an early fatality within one mile and individual risk of a cancer fatality within ten

-7-



SPAR HRA Human Error Worksheet (Page 1 of 3)

Plant:, Initiating Event:, Sequence Number:, Basic Event Code:

Basic Event Context:
Basic Event Description:,

Does this task contain a significant amount of diagnosis activity? YES (start with PartI,p.1) NO (skip Part I, p. 1; start with Part Il, p. 2)
‘Why?,

Part |. DIAGNOSIS
A. Evaluate PSFs for the diagnosis portion of the task.

PSFs PSF Levels ‘Multiplier for If non-nominal PSF fevels are selected, please note
Diagnosis specific reasons in this column
Available Time Inadequate time P(failure) = 1.0

Barely adequate time <20 m
‘Nominal time . 30min
“Extra time 560 min
Expansive time >24 hrs.
Stress Extreme
igh
‘Nominal
‘Complexity ~_Highly complex

“Moderately compiex
Nominal
‘Obvious diagnosis
Experience/Training ‘Low
Nominal
igh
Procedures Not available
“Available, bi

6
“Nominal i
ig .
Fitness for Duty Unfit P(failure) = 1.0
‘Degraded Fitness 3
‘Nominal i

Work Processes.

B. Calculate the Diagnosis Failure Probability

(1) If all PSF ratings are nominal, then the Diagnosis Failure Probability = 1E-2

(2) Otherwise, Time Stress Complexity Experience/_ Procedures Ergonomics Fitness Work
Training for Duty Processes
Diagnosis: 1E-2x__ x__ X__ x x x x x =
Diagnosis

Failure Probability


Gary M. Holahan 4

References: 1. Memorandum from G. Holahan to T. King dated March 26, 1999
2. Memorandum from A. Thadani to S. Collins dated November 12, 1999
3. Memorandum from F. Eltawila to G. Holahan dated August 25, 2000
4. Code Manual for MACCS2, NUREG/CR-6613, May 1998
5. Memorandum from R. Barrett to J. Flack dated August 25, 2000
6. Memorandum from S. Collins to A. Thadani dated September 11, 2000
7. Accident Source Terms for Light-Water Nuclear Power Plants, NUREG-1465,
February 1995
8. Chernobyl Ten Years On, Radiological and Health Impact, An Appraisal by
the NEA Committee on Radiation Protection and Public Health, November
1995
cc: T. Collins
R. Barrett
J. Hannon
J. Wermiel

G. Hubbard


benefit (and significant analytical effort) if the potential additional 2 days until the zirconium fire
began were added to the timing.

After the preliminary draft risk assessment was released in June 1999, the staff sent the
assessment to INEEL for review and held public meetings and a workshop to assure that
models appropriately accounted for the way decommissioning plants operate today and to help
determine if some of the assumptions we made in the preliminary draft risk assessment needed
improvement. Following a workshop, NEI provided a list of general commitments (see
Appendix 5) that proved instrumental in refining the assumptions and models in the draft final
tisk assessment. Working with several PRA experts, the staff subsequently developed
improved HRA estimates for events that lasted for extended periods.

This appendix describes how the risk assessment was performed for beyond design bases
internal event accident sequences (i.e., sequences of equipment failures or operator errors that
could lead to a zirconium cladding fire and release of radionuclides offsite). Event trees and
fault trees were developed that model the initiating events and system or component failures
that lead to fuel uncovery (these trees are provided in Appendix 2A).

Appendix 2 A2-3


4.5.8 Top Event OFD - Recovery From Offsite Sources
4.5.8.1 Event Description and Timing

Given the failure of recovery actions using onsite sources, this event accounts for recovery of
coolant makeup using offsite sources such as procurement of a fire engine. This event is
represented by the fault trees LOI-OFD-LGLK, LOI-OFD-SMLK and LO!-OFD-SMLK-L for the
large break and two small break scenarios, respectively.

4.5.8.2 Relevant Assumptions

. The operator has 40 to 65 hours depending on the break size to provide makeup inventory
and cooling.

. Procedure explicitly states that if the water level drops below a certain level (e.g., 15 ft
below normal level) operator must initiate recovery using offsite sources.

. Operator has received formal training and there are procedures to guide him.
. Offsite resources are familiar with the facility.

4.5.8.3 Quantification

Human Error Probabilities

The only new basic events in these functional fault trees are HEP-INV-OFFST-LK and HEP-
INV-OFFST. They were quantified using SPAR HRA worksheets. The diagnosis of the need to
initiate the action is considered totally dependent on the recognition of the need to initiate
inventory makeup with the fire water system. The PSFs are as follows: extreme stress (it’s the
last opportunity for success), high complexity because of the involvement of offsite personnel,
highly trained staff with good procedures, good ergonomics (equipment is available to make
offsite support straightforward) and good work processes. For both cases, a low level of
dependence was assumed on the failure of prior tasks.

4.5.8.4 Basic Event Probabilities

Basic Event Basic Event Probability

HEP-INV-OFFST-LK 5.0E-2

HEP-INV-OFFSITE 5.0E-2
4.5.9 Summary

Table 4.5 presents a summary of basic events.

As in the previous cases, the frequency of fuel uncovery can be seen to be very low. Again, a
careful and thorough adherence to NEI commitments 2, 4, 5, 8 and 10, the assumption that

Appendix 2A A2A-63


using fault tree LOC-OFB, for the upper two branches, and LOC-OFB-L for the lowest branch.
Note that in this fault tree event HEP-INV-OFFSITE is ORed with the failure of the operator to
recognize the need to start the firewater system (event HEP-RECG-FWSTART or
HEP-RECG-FWSTART-L , described in Section 4.1.5.3). In essence, if the operators fail to
recognize the need for firewater, it is assumed they will fail to recognize the need for other
offsite sources of make-up.

4.1.6.2 Relevant Assumptions

. The operators have 88 hours to provide makeup and inventory cooling.

. Procedures and training are in place that ensure that offsite resources can be brought to
bear (NEI commitment no. 2 and 4), and that preparation for this contingency is made

when it is realized that it may be necessary to supplement the pool make-up.

. Procedures explicitly state that if the water level drops below a certain level (e.g., 15 ft
below normal level) operator must initiate recovery using offsite sources.

° Operators have received formal training in the procedures.
. Offsite resources are familiar with the facility.

4.1.6.3 Quantification

Human Error Probabilities

The event HEP-INV-OFFSITE represents failure to recognize that it is necessary to take the
extreme measure of using offsite sources, given that even though there has been ample time up
to this point to attempt recovery of both the SFP cooling system and both firewater pumps it has
not been successful. This top event should include contributions from failure of both the
diagnosis of the need to provide inventory from offsite sources, and of the action itself. The
availability of offsite resources is assumed not to be limiting on the assumption of an expansive
preparation time. However, rather than use a calculated HEP directly, a low level of
dependence on the failure to recognize the need to initiate the firewater system was assumed.

4.1.6.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-INV-OFFSITE 5.0E-2
4.1.7 Summary

Table 4.1 presents a summary of basic event probabilities used in the event tree quantification.

Based on the assumptions made, the frequency of fuel uncovery can be seen to be very low. A
careful and thorough adherence to NEI commitments 2, 5, 8 and 10 is crucial to establishing the
low frequency. In addition, however, the assumption that walk-downs are performed on a

Appendix 2A A2A-22


Figure A2e-1

Annual Average Number of Tornadoes per
10,000 Square Miles by State, 1950-1995

Figure Average Annual Number of Strong-Violent (F2-F5) A2e-2
Tornadoes per 10,000 Square Miles by State

Appendix 2E "A2E-9


by a prolonged graphite fire) than an SFP accident. In subsequent discussions, this source term
is referred to as the high ruthenium source term.

The consequences obtained using the source term in NUREG-1465 (which treats ruthenium as
a less volatile fission product) in conjunction with SFP fission product inventories are provided in
Table 3.7-2 for comparison. In subsequent discussions, this source term is referred to as the
low ruthenium source term.

The consequence calculations for both the high and low ruthenium source terms assume that all
of the fuel assemblies discharged in the final core off-load and the previous 10 refueling outages
participate in the SFP fire. These assemblies are equivalent to about 3.5 reactor cores.
Approximately 85 percent of all the ruthenium in the pool is in the last core off-loaded since the
ruthenium-106 half-life is about 1 year. For cesium-137, with a 30-year half-life, the inventory
decays very slowly and is abundant in all of the batches considered. The staff assumed that the
number of fuel assemblies participating in the SFP fire remains constant and did not consider
the possibility that fewer assemblies might be involved in an SFP fire in later years because of
substantially lower decay heat in the older assemblies. Based on the limited analyses
performed to date, fire propagation is expected to be limited to less than two full cores 1 year
after shutdown (see Appendix 1A). Thus, the assumption that 3.5 cores participate adds some
conservatism to the calculation of long-terms effects associated with cesium, but is not important
with regard to the effects of ruthenium.

The results for early fatality and societal dose (person-rem) consequences for an SFP accident
are graphed in Figures 3.7-1 and 3.7-2. The early fatality plots are truncated at a value of one
early fatality since fractions of a fatality are not meaningful. Since no early fatalities were
predicted for the low ruthenium source term with early evacuation, a curve is not shown for that
case in Figure 3.7-1. Because latent cancer fatalities are directly proportional to societal dose
through a dose-to-cancer-risk conversion factor within the MACCS2 consequence code (Ref. 9),
results for latent cancer fatalities are not displayed separately.

3-31


MR
MW
MWD
MTU

NEI
NRC
NRR

OSC

POE
POF
PPG

PWR

QA
QHO

RES
RG

SDA

SFP
SFPC
SFPCC
SHARP
SNL
SRM
SRP
SSC
SSE

TS
TSC

UKAEA

WIPP

Maintenance Rule
megawatt
megawatt-day
metric ton uranium

Nuclear Energy Institute
Nuclear Regulatory Commission
NRC Office of Nuclear Reactor Regulation

operations support center

probability of exceedance
probability of failure

pool performance guideline
probabilistic risk assessment
pressurized-water reactor

quality assurance
quantitative health objective

NRC Office of Regulatory Research
regulatory guide

staff decommissioning assumption

spent fuel

spent fuel pool

spent fuel pool cooling system

spent fuel pool cooling and cleaning system
Spent Fuel Heatup Analytical Response Program
Sandia National Laboratory

staff requirements memorandum

standard review plan

systems, structures, and components

safe shutdown earthquake

technical specification
technical support center

United Kingdom Atomic Energy Authority

waste isolation pilot plant

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A4D-9

Appendix 4D


Technical Study of Spent Fuel Pool Accidents
at Decommissioning Plants

Table of Contents

ABSTRACT... 0.20. e tenet eect et eee tenes iii
EXECUTIVE SUMMARY ..... 00... tenet bene eenee ix
ACKNOWLEDGMENTS ........... 02.0. c bocce cece eet eee eee e eens xv
1.0 INTRODUCTION ............ 056 c ccc cece cece ence eee eee e eee 1-1
2.0 THERMAL-HYDRAULIC ANALYSES ........0.0.0. 0000 c cece eect e tenet eee 2-1

3.0 RISK ASSESSMENT OF SPENT FUEL POOLS AT DECOMMISSIONING PLANTS . . 3-1
3.1 Basis and Findings of SFP Risk Assessment ...............00--.00-002 eee 3-2
3.2 Characteristics of SFP Design and Operations for a Decommissioning Plant .. . . 3-3
3.3 Estimated Frequencies of Spent Fuel Uncovery and Assumptions That Influence

the Results 20.0... eee teen ete ent ee eeee 3-6
3.3.1 Internal and External Initiator Frequency of Spent Fuel Pool Uncovery... 3-7
3.3.2 Important Assumptions ........... 0.000.
3.4 Internal Event Scenarios Leading to Fuel Uncovery .
3.4.1 Loss of Cooling ................ 0.00
3.4.2 Loss of Coolant Inventory ...........00. 0.00. e cence eee
3.4.3 Loss of Offsite Power from Plant-Centered and Grid Related Events ... 3-15
3.4.4 Loss of Offsite Power from Severe Weather Events................. 3-15
3.4.5 Internal Fire ........ 0. eee eee eee 3-16
3.4.6 Heavy Load Drops ......... 0... e ccc ere ees 3-16
3.4.7 Spent Fuel Poo! Uncovery Frequency at Times Other Than 1 year After
Shutdown .... 0.2... cece eee eee 3-18
3.5 Beyond Design Basis Spent Fuel Pool Accident Scenarios (External Events) . . 3-18
3.5.1 Seismic Events........0.0 20000000 occ eee eee eee 3-18
3.5.2 Aircraft Crashes . 3-23
3.5.3 Tornadoes ........ : 3-24

3.6 Criticality in Spent Fuel Pool ............
3.7 Consequences and Risks of SFP Accidents .
3.7.1 Consequences of SFP Accidents ...

3.7.2 Risk Modeling for SFP Accidents . : 3-34
3.7.3 Risk Results... 20.2.0. 3-39

4.0 IMPLICATIONS OF SPENT FUEL POOL (SFP) RISK FOR REGULATORY
REQUIREMENTS ...........0.0 0.00000 e cece cece cece ee A
4.1 Risk-informed Decision Making . . 4-1
4.1.1. Increases in Risk ..... » 4-2
4.1.2 Defense-in-Depth 2.0.0... 0000. 4-6


=csas26 parm=size=10000000
KENO-VI Input for Storage Cell Calc. High Density Poisoned Rack
238groupndf5 latticecell

‘Data From SAS2H - Burned 5 w/o Fuel
0-16 = 1 0 0.4646E-01 300.00 end
kr-83 1 0 0.3694E-05 300.00 end
th-103 1 0 0.2639E-04 300.00 end
th-105 1 0.0.6651E-07 300.00 end
ag-109 10 0.4459E-05 300.00 end
xe-131 10 0.2215E-04 300.00 end
'xe-135 1 0 0.9315E-08 300.00 end
cs-133 10 0.5911E-04 300.00 end
cs-134 10 0.5951E-05 300.00 end
es-135 100.2129E-04 300.00 end
ba-140 1 00.1097E-05 300.00 end
la-140 1 0.0.1485E-06 300.00 end
nd-143 1 0 0.4070E-04 300.00 end
nd-145 1 0 0.3325E-04 300.00 end
pm-147 1 0 0.8045E-05 300.00 end
pm-148 1 0 0.4711E-07 300.00 end
pm-148 1 0 0.6040E-07 300.00 end
pm-149 1 0 0.6407E-07 300.00 end
sm-147 10 0.3349E-05 300.00 end
sm-149 1 0 0.1276E-06 300.00 end
sm-150 10 0.1409E-04 300.00 end
sm-151 1 0. 0.7151E-06 300.00 end
sm-152 1 0 0.5350E-05 300.00 end
eu-153 1 0 0.4698E-05 300.00 end
eu-154 1 0.0.1710E-05 300.00 end
eu-155 1 0 0.6732E-06 300.00 end
gd-154 100.1215E-06 300.00 end
gd-155 1 00.5101E-08 300.00 end
gd-156 10 0.2252E-05 300.00 end
gd-157 10 0.3928E-08 300.00 end
gd-158 10 0.6153E-06 300.00 end
gd-160 100.3549E-07 300.00 end
u-234 10 0.6189E-07 300.00 end
u-235 = 1 0 0.3502E-03 300.00 end
u-236 10 0.1428E-03 300.00 end
u-238 10 0.2146E-01 300.00 end
np-237 10 0.1383E-04 300.00 end
pu-238 1 0 0.4534E-05 300.00 end
pu-239 10 0.1373E-03 300.00 end
pu-240 10 0.5351E-04 300.00 end
pu-241 1 0 0.3208E-04 300.00 end
pu-242 10 0.1127E-04 300.00 end
am-241 10 0.9976E-06 300.00 end
am-242 1 0 0.2071E-07 300.00 end
am-243 1 0 0.2359E-05 300.00 end

Appendix 3 A3-11


well. Therefore this value consists of two errors. The diagnosis error was calculated using
SPAR, and the walkdown error was calculated using THERP. The relevant performance
shaping factors included greater than 24 hours for diagnosis, extreme stress, moderately
complex task (because of potential complications from severe weather), diagnostic procedures,
and good work processes. A low dependence was applied to the walk-down error.

Event HEP-SFP-STR-LP2 represents operator failure to restart/realign the SFP cooling system
in 9 hours. The operators can restart the previously running pump and may not have to make
any valve alignment. If they decide to restart the standby pump they may have to make some
valve alignment. This error was quantified using SPAR. The relevant performance shaping
factors included expansive time, extreme stress because of severe weather, moderately
complex task because of potential valve lineups and severe weather, poor ergonomics because
of severe weather, and good work process.

If the system fails to start and run for a few hours then the operators would try to get the system
repaired. Assuming that it takes another two shifts (16 hours) to contact maintenance
personnel, make a diagnosis, and get new parts, and assuming an average repair time of

40 hours, there is not sufficient time to fix the system. Therefore, no credit was given for repair
of the SFP cooling system.

Non-HEP Probabilities

Fault tree LP2-OCS represents failure of the SFP cooling system to restart and run. Hardware
failure rates have been taken from INEL-96/0334. It is assumed that the SFPC system will be
maintained since it is required to be running all the time.

4.4.3.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-DIAG-SFPLP2 1.0E-5
HEP-SFP-STR-LP2 5.0E-4
SPC-CKV-CCF-H 1.9E-5
SPC-CKV-CCF-M 3.2E-5
SPC-HTX-CCF 1.9E-5
SPC-HTX-FTR 2.4E-4
SPC-HTX-PLG 2.2E-5
SPC-PMP-CCF 5.9E-4
SPC-PMP-FTF-1 3.9E-3
SPC-PMP-FTF-2 3.9E-3

Appendix 2A A2A-46


Table A4-13 Comparison of Analysis Assumptions

Parameter NUREG/CR- | NUREG/CR-6451 | Case 3 Case 4
4982 (BWR) (BWR)
population 100 0-30 mi: 1000 100 100
density 30-50 mi: 2300
(persons/ (city of 10 million
mile’) people, 280
outside of city)
50-500 mi: 200
meteorology | uniform wind representative for | Surry Surry
rose, average | continental U.S.
weather
conditions
radionuclide 11 batches of | full fuel pool after {| 11 batches of 11 batches of
inventory spent fuel decommissioning | spent fuel, spent fuel plus
(3300 increased by last of rest core,
assemblies) x1.7 increased by x1.7
exclusion not reported Ami none none
area
emergency relocation at relocation at one NUREG-1150 NUREG-1150
response one day if day if projected Surry analysis Surry analysis
projected doses exceed 25 | (see above) (see above)
doses exceed | rem
25 rem

Table A4-14 Comparison with NUREG/CR-4982 Results

Decay Time in Distance Societal Dose (person-Sv)
Spent Fuel (miles)
Pool NUREG/CR- Case 3 Case 3 Rerun
4982
30 days 0-50 26,000 20,900 16,700
0-500 710,000 449,000 379,000
90 days 0-50 26,000 20,400 16,500

The NUREG/CR-6451 results included consequence estimates for societal dose, cancer
fatalities, and prompt fatalities for a decommissioned reactor for a severe SFP accident
occurring 12 days after the final shutdown. The Case 4 results for 30 days after final shutdown
were compared against the NUREG/CR-6451 results, because (1) they included the entire last
core in the SFP and (2) Case 4 had a uniform population density which could be easily
adjusted to approximate that in NUREG/CR-6451. Differences between Case 4 and

Appendix 4

A4-14



different source terms to show the impact of source term uncertainties on results. In

Section 3.7.2, the severe accident consequences for either the early or late evacuation cases
are assigned to each of the major types of SFP accidents, as appropriate, and then combined
with the respective event frequencies to provide a scoping estimate of SFP risks. The risks of
SFP accidents are shown to meet the Commission's safety goals. The impact of changes in EP
regulations on these risk measures is discussed later in Section 4.

3.7.1. Consequences of SFP Accidents

Earlier analyses in NUREG/CR-4982, “Severe Accidents in Spent Fuel Pools in Support of
Generic Issue 82,” and NUREG/CR-6451, “A Safety and Regulatory Assessment of Generic
BWR and PWR Permanently Shutdown Nuclear Power Plants,” included a limited analysis of
the offsite consequences of a severe SFP accident occurring up to 90 days after the last
discharge of spent fuel into the SFP. The analysis showed that the consequences of an SFP
accident could be comparable to those for a severe reactor accident. As part of its effort to
develop generic, risk-informed requirements for decommissioning, the staff performed a further
analysis of the offsite radiological consequences of beyond-design-basis SFP accidents.
Varying the evaluation and other modeling assumptions, the staff performed an initial set of
calculations to extend the earlier analyses to SFP accidents occurring 1 year after plant
shutdown and to supplement the earlier analyses with additional sensitivity studies. The results
of these calculations were documented in the February 2000 study, and are provided in
Appendix 4.

Subsequently, the ACRS raised issues with the source term and plume modeling for SFP
accidents. In particular, the ACRS believed that the ruthenium and fuel fines releases were too
low and the plume was too narrow. To address these issues, the staff performed additional
sensitivity studies, as documented in Appendix 4A of this study.

To provide insight into the impact on results of decay times shorter or longer than 1 year,
additional consequence calculations were performed using fission product inventories at 30 and
90 days and 2, 5, and 10 years after final shutdown. The results of these consequence
calculations were used as the basis for assessing the risk from SFP accidents. These results
are summarized in Tables 3.7-1 and 3.7-2 for several key consequence measures, and are
described in more detail in Appendix 4B. These consequences are conditional upon the
occurrence of an accident that results in an SFP fire, i.e., the consequences are on a “per event”
rather than a “per year,” basis and do not account for the probability of the event.

These calculations were based on the Surry site, although the SFP accident consequences
could be greater at higher population sites, the quantitative health objectives used in
comparisons to the Commission's Safety Goals (see Section 3.7.3) represent risk to the average
individual within 1 mile and 10 miles of the plant, and should be relatively insensitive to the site
specific population.

3-28


. Training plans are revised as needed to reflect the changes in equipment configuration as
they occur

4.5.4.3 Quantification
Human Error Probabilities

The top event LOI-IND, for small leaks, includes two HEPs, depending on whether the control
room alarms have failed, or the operators failed to respond to the alarms. If the operators failed
to respond to control room alarms, then event HEP-WLKDWN-DEPEN models the failure of the
next shift to recognize the loss of cooling during a walkdown or during a control room review,
taking into account a potential dependence on event HEP-DIAG-ALARM. A low dependence is
assumed. If the alarms failed, then event HEP-WWLKDWN-LOI models operator's failure to
recognize the loss of inventory during walk-downs, with no dependence on previous HEPs.
Because only one crew is credited, the HEP is estimated as 5E-03.

This failure probability is developed using THERP, and is based upon three individual failures:
failure to carry out an inspection, missing a step in a written procedure, and misreading a
measuring device.

The top event LOI-IND-L is modeled taking into account several opportunities for recovery by
consecutive crews, and because the indications are so compelling no dependency is assumed
between this HEP and the prior event.

4.5.4.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-WLKDWN-DEPEN 5.0E-2
HEP-WLKDWN-LOI-L 1,0E-5
HEP-WLKDWN-LOI! 5.0E-3

4.5.5 Top Event OIS - Operator Isolates Leak and Initiates SFP Make-up
4.5.5.1 Event Description and Timing

This top event represents the operator's failure to isolate a large leak and initiate the SFP
makeup system before the pool level drops below the SFP cooling system suction, and is
represented by the fault tree LOI-OIS-U. Failure requires that the operators must provide the
inventory using the firewater system or offsite resources.

The critical action is the isolation of the leak. With the leak size assumed, and on the
assumption that the low level alarm is set at 1 foot below the normal level, the operators have 4
hours to isolate the leak. Once the leak has been isolated, there would be considerable time
available to initiate the normal make-up, since pool heat up to the point of initiation of boiling
takes several hours.

Appendix 2A A2A-57


examine the impact of the proposed changes continues to reflect the actual reliability and
availability of SSCs that have been evaluated. This will ensure that the conclusions that have
been drawn will remain valid.

Applying this guideline for the SFP risk evaluation results in identification of four primary areas
for performance monitoring: (1) the performance and reliability of SFP cooling and associated
power and inventory makeup systems, (2) the Boraflex condition for high-density fuel racks,

(3) crane operation and load path control for cask movements, and (4) onsite emergency
response capabilities. The following monitoring should continue after decommissioning in order
to assure SFP risk remains low:

+ Performance and reliability monitoring of the SFP systems, heat removal, AC power, and
inventory should comply with the provisions of the Maintenance Rule (10 CFR 50.65).

+ The current monitoring programs identified in licensee’s responses to Generic Letter 96-04
(Ref. 2) with respect to monitoring of the Boraflex absorber material should be maintained
by decommissioning plants until all fuel is removed from the SFP. This staff assumption is
stated in SDA #7 (see Table 4.1-2).

+ Heavy load activities and load paths should be monitored and controlled by the licensee in
accordance with IDC #1 (see Table 4.1-1).

+ Licensees should continue to provide a level of onsite capabilities to assure prompt
notification of offsite authorities, characterization of potential releases, development of
protective action recommendations, and communication with the public. These capabilities
should be monitored by holding periodic onsite exercises and drills.

The staff concludes that continued compliance with the Maintenance Rule, the IDCs, and the
SDAs, together with some level of EP, provides a reasonable level of monitoring of SFP safety.

4-10


sabotage. However, any revisions in security provisions should be constrained by an
effectiveness assessment of the safeguards provisions against a design-basis threat. Because
the possibility of a zirconium fire leading to a large fission product release cannot be ruled out
even many years after final shutdown, the safeguards provisions at decommissioning plants
should undergo further review. The results of this study may have implications on previous
exemptions at decommissioning sites, devitalization of spent fuel pools at operating reactors
and related regulatory activities.

The staff's risk analyses were complicated by a lack of data on severe-earthquake return
frequencies, source term generation in an air environment, and SFP design variability. Although
the staff believes that decommissioning rulemaking can proceed on the basis of the current
assessment, more research may be useful to reduce uncertainties and to provide insights on
operating reactor safety. In particular, the staff believes that research may be useful on source
term generation in air, which could also be important to the risk of accidents at operating
reactors during shutdowns, when the reactor coolant system and the primary containment may
both be open.

In summary, the study finds that:

1. The risk at decommissioning plants is low and well within the Commission's safety goals.
The risk is low because of the very low likelihood of a zirconium fire even though the
consequences from a zirconium fire could be serious.

2. The overall low risk in conjunction with important differences in dominant sequences
relative to operating reactors, results in a small change in risk at decommissioning plants if
offsite emergency planning is relaxed. The change is consistent with staff guidelines for
small increases in risk.

3. Insurance, security, and EP requirement revisions need to be considered in light of other
policy considerations because a criterion of “sufficient cooling to preclude a fire” cannot be
satisfied on a generic basis.

4. Research on source term generation in an air environment would be useful for reducing
uncertainties.

5-3


NUREG-1738

Technical Study of Spent
Fuel Pool Accident Risk
at Decommissioning

OP Nuclear Power Plants

U.S. Nuclear Regulatory Commission ras
Office Nuclear Reactor Regulation z eA 3
Washington, DC 20555-0001 [Nets

teat

x
y
&



should be low. The design features include the physical plant characteristics (e.g., nature and
number of alarms, available mitigation equipment) and the operational practices include
operational and management practices (including crew structure and individual responsibilities),
procedures, contingency plans, and training. Since the details will vary from plant to plant, the
focus is on general design features and operational practices that can support low non-response
probabilities.

Section 3.2.2 discusses the differences between the full power and decommissioning modes of
operation as they impact human reliability analysis, and the issues that need to be addressed in
the analysis of the decommissioning mode are identified. Section 3.2.3 discusses the factors
that recent studies have shown to be significant in establishing adequacy of human
performance.

3.2.2 Analysis Approach

The HRA approaches that have been developed over the past few years have primarily been for
use in PRAs of nuclear power plants at full power. Methods have been developed for assessing
the likelihood of errors associated with routine processes such as restoration of systems to
operation following maintenance, and those errors in responding to plant transients or accidents
from full power. For SFP operation during the decommissioning phase, there are unique
conditions not typical of those found during full-power operation. Thus the human reliability
methods developed for full power operation PRAs, and their associated error probabilities, are
not directly applicable. However, some of the methods can be adapted to provide insights into
the likelihood of failures in operator performance for the SFP analysis by accommodating the
differences in conditions that might impact operating crew performance in the full power and
decommissioning phases. There are both positive and negative aspects of the difference in
conditions with respect to the reliability of human performance.

Examples of the positive aspects are:

. For most scenarios, the time-scale for changes to plant condition to become significant are
protracted. This is in contrast to full power transients or accidents in which response is
required in a relatively short time, ranging from a few minutes to a few hours. In the staffs
analysis, times ranging from 100 to greater than 220 hours were estimated for heat up and
boil off following loss of SFP cooling. Thus, there are many opportunities for different plant
personnel to recognize off-normal conditions, and a long time to take corrective action,
such as making repairs, hooking up alternate cooling or inventory makeup systems, or
even bringing in help from offsite.

° There is only one function to be maintained, namely decay heat removal, and the systems
available to perform this function are relatively simple. By contrast, in the full power case
there are several functions that have to be maintained, including criticality, pressure
control, heat removal, containment integrity.

° With respect to the last point, it is also expected that the number of controls and indications
that are required in the control room are considerably fewer than for an operating plant,
and therefore, there is less cause for confusion or distraction.

Appendix 2A A2A-4


NONRECOVERY
OF OSP FROM PLANT
‘CENTERED EVENTS

NONRECOVERY
(OF OS? FROM PLANT
‘CENTERED EVENTS

REGOSRPC

Pec
weg

A

eeRAMSTS

RES

ver Bese,

a oO

Oo 2D oO

Oa
Oe

=

A-9

Q 2



Hardware Failure Probabilities

The value used for failure of the detectors, SFP-FIRE-DETECT (5.0E-3), was taken from
OREDA-92 (Ref. 14). The value used for local electrical faults leading to alarm channel failure,
SFP-FIRE-LOA (2.0E-3), was estimated based on information in reference 13.

4.2.2.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-DIAG-ALARM 3.0E-4
SFP-FIRE-LOA 2.0E-3
SFP-FIRE-DETECT 5.0E-3

4.2.3 Top Event IND - Other Indications of Loss of Cooling
4.2.3.1 Event Description and Timing

This event models the failure of the operators to recognize the loss of SFP cooling resulting
from a fire, given that either the fire alarm system failed or was not attended to. Since the
assumed consequences of not attending to the alarm are a fire large enough to cause loss of
power to the facility, the indications available to the operator during a walk-down include clear
effects of the fire, both from visible evidence and the smell of burning, as well as the lack of
power. Ultimately, if no action is taken to restore cooling, the high area temperature and
humidity, and low water level from boiloff will become increasingly evident. The operators have
more than 10 shifts (about 131 hours) to discover the loss of SFP cooling. Success for this
event is defined as the operators recognizing the abnormal condition and understanding the
need to take action within this time. This event is modeled by fault tree FIR-IND.

4.2.3.2 Relevant Assumptions

. Operators perform walk-downs once per shift (every 8 to 12 hours) and walk-downs are
required to be logged.

. If the fire is discovered during the walk-down, the SFPC system is assumed to be
damaged to the extent where repair will not be feasible within a few days.

. Local instrumentation and alarms are destroyed in a fire which is not extinguished within
20 minutes.

° Procedures are available to guide plant operators for off-normal conditions, and operators
are trained in these procedures (NEI commitment no. 2).

Appendix 2A A2A-27


. Regular test and maintenance is performed on instrumentation (NEI commitment no. 10).

¢ Procedures are available to guide the operators on response to off-normal conditions, and
the operators are trained on the use of these procedures (NEI commitment no. 2).

. System drawings are revised as needed to reflect current plant configuration.
. SFP water level indicator is provided in the control room (NEI commitment no. 5).

. SFP low-water level alarm (narrow range) is provided in the control room (NEI commitment
no. 5).

. Low level alarm set point is set to one foot below the normal level.

4.5.3.3 Quantification

Human Error Probabilities

One operator error, HEP-DIAG-ALARM is modeled under this top event. This event represents
operator failure to respond after receiving a low-level alarm. Success is defined as the operator
investigating the alarm and identifying the cause. This failure was quantified using The

Technique for Human Error Prediction (THERP) Table 20-23. No distinction is made between
the two leak sizes because this is treated as a simple annunciator response.

Non-HEP Probabilities

The value used for local faults leading to alarm channel failure, SPC-LVL-LOP (2.0E-3), was
estimated based on information in NUREG-1275, Volume 12. This includes both local electrical
faults and instrumentation faults.

4.5.3.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-DIAG-ALARM 3.0E-4
SPC-LVL-LOP. 2.0E-3

4.5.4 Top Event IND - Other Indications of Inventory Loss
4.5.4.1 Event Description and Timing

This top event models operator failure to recognize the loss of inventory during walk-downs over -
subsequent shifts. Indications available to the operators include read-outs in the control room,
and a visibly decreasing water level. Eventually, when pool cooling is lost the environment

would become noticeably hot and humid. Success for this event, in the context of the event

tree, is treated differently for the small and large leaks.

Appendix 2A A2A-55


ABSTRACT

This study contains the results of the NRC staff's evaluation of the potential accident risk ina
spent fuel pool at decommissioning plants in the United States. This study was prepared to
provide a technical basis for decommissioning rulemaking for permanently shutdown nuclear
power plants. This study describes a modeling approach of a typical decommissioning plant
with design assumptions and industry commitments; the thermal-hydraulic analyses performed
to evaluate the behavior of spent fuel stored in the spent fuel pool at decommissioning plants;
the risk assessment of spent fuel pool accidents; the consequence calculations; and the
sensitivity study and implications for decommissioning regulatory requirements. Preliminary
drafts of this study were issued for public comments and technical reviews in June 1999 and
February 2000. Comments from interested stakeholders, the Advisory Committee on Reactor
Safeguards, and other technical reviewers have been taken into account in preparing this study.
A broad quality review was also carried out at the Idaho National Engineering and Environment
Laboratory, and a panel of human reliability analysis experts evaluated the report’s
assumptions, methods, and modeling. Public comments on draft versions of this study are
discussed in Appendix 6 of this NUREG.


FAILURE TO INMATE
NORMAL MAKEUP
(SMALL LEAK)
tor Lt
; ;
OPERATOR FAILS TO.
KEYS, Sanrenee
FAILS (SMALL LEAK)
SFP-REGMKUP-F HEP-MKUP-START-E
1
—
aan

LEAK

= 4
aman es ry
INITIATE MAKEUP BY USE NEED TO START AW FUME
EWEN oy PONPS: (PLECTRIC OR DISSE)
pay cey Q.
Fag Daeg DIESEL AND ELECTRIC
rence FALTOSTART ND RUN

ben



Attachment 2

Comments Concerning Seismic Screening
And Seismic Risk of Spent Fuel Pools for
Decommissioning Plants
by
Robert P. Kennedy
October 1999

prepared for
Brookhaven National Laboratory
2. Introduction

I have been requested by Brookhaven National Laboratory, in support of the Engineering
Research Applications Branch of the Nuclear Regulatory Commission, to review and comment
on certain seismic related aspects of References 1 through 4. Specifically, I was requested to
comment on the applicability of using seismic walkdowns and drawing reviews conducted
following the guidance provided by seismic screening tables (seismic check lists) to assess that
the risk of seismic-induced spent fuel pool accidents is adequately low. The desire is to use these
seismic walkdowns and drawing reviews in lieu of more rigorous and much more costly seismic
fragility evaluations. It is my understanding that the primary concern is with a sufficiently gross
failure of the spent fuel pool so that water is rapidly drained resulting in the fuel becoming
uncovered. However, there may also be a concern that the spent fuel racks maintain an
acceptable geometry. It is also my understanding that any seismic walkdown assessment should
be capable of providing reasonable assurance that seismic risk of a gross failure of the spent fuel
pool to contain water is less than the low 10 mean annual frequency range. My review
comments are based upon these understandings.

2. Background Information

The NRC Draft Technical Study of Spent Fuel Pool Accidents (Ref. 1) assumes that
spent fuel pools are seismically robust. Furthermore, it is assumed that High-Confidence-Low-
Probability-of Failure (HCLPF) seismic capacity of these pools is in the range of 0.4 to 0.5g
peak ground acceleration (PGA). This HCLPF capacity (Cyc: pr) corresponds to approximately a
1% mean conditional probability of failure capacity (C,.,), i.e.:

Cucupr ® Cie, (1)
as shown in Ref. 10.
In Ref. 5, detailed seismic fragility assessments have been conducted on the gross

structural failure of spent fuel pools for two plants: Vermont Yankee (BWR), and Robinson
(PWR). The following HCLPF seismic capacities are obtained from the fragility information in

A2B-11


Examples of the negative aspects are:

The plant operation is not as constrained by regulatory tools (technical specifications are
not as comprehensive and restrictive as they are for operating plants), and there is no
requirement for emergency procedures.

Because the back-up systems are not automatically initiated, operator action is essential to
successfully respond to failures of the cooling function.

There is expected to be little or no redundancy in the onsite mitigating capability as
compared with the operating plant mode of operation. (In the staff's initial evaluation,
because little redundant onsite equipment was assumed to be available, the failure to bring
on offsite equipment was one of the most important contributors.) This implies that repair
of failed functions is relatively more significant in the risk analysis for the SFP case.

In choosing an approach for developing the estimates documented in this study, the following
issues were considered to be important:

.

Because of the long time scales, it is essential to address the potential for recovery of
failures on the part of one crew or individual by other plant staff, including subsequent
shifts.

Potential sources of dependency that could lead to a failure of the organization as a whole
to respond adequately should be taken into account.

The approach should be consistent with current understanding of human performance
issues (Refs. 7, 8, and 9).

Those factors that the industry has suggested that will help ensure adequate response
(instrumentation, monitoring strategies, procedures, contingency plans) should be
addressed (Ref. 4).

Where possible, any evaluations of human error probabilities (HEPs) should be calibrated
against currently acceptable ranges for HEPs.

The reasoning behind the assumptions made should be transparent.

3.2.3 Human Performance Issues

In order to be successful in coping with an incident at the facility, there are three basic functions
that are required of the operating staff, and these are either explicit (awareness) or implicit
(situation assessment and response planning and response implementation) in the definitions of
the human failure events in the PRA model.

Piant personnel must be able to detect and recognize when the spent fuel cooling function
is deteriorating or pool inventory is being lost (Awareness).

Appendix 2A A2A-5


important in the analysis for increasing the likelihood that offsite organization can respond
effectively.

3.4.5 Internal Fire

This event tree models the loss of SFP cooling caused by internal fires. The staff assumed that
there is no automatic fire suppression system for the SFP cooling area. The fuel handler may
initially attempt to manually suppress the fire if the fuel handler responds to the control room or
local area alarms. If the fuel handler fails to respond to the alarm or is unsuccessful in
extinguishing the fire within the first 20 minutes, the staff assumes that the SFP cooling system
will be significantly damaged and cannot be repaired. Once the inventory level drops below the
SFP cooling system suction level, the fuel handlers have about 85 hours to provide some sort of
alternative makeup, either by using the site firewater system or by calling upon offsite resources.
The staff assumes that the fire damages the plant power supply system and the electrical
firewater pump is not available.

The calculated fuel uncovery frequency for this event is 2.3x10* per year. As in the previous
cases, this estimate was based on IDCs #2, #5, #8, and #10 and on SDA #2 and SDA #3. In
addition, IDC #3, the commitment to have procedures in place for communications between
onsite and offsite organizations during severe weather, is also important in the analysis for
increasing the likelihood that offsite organizations can respond effectively to a fire event
because the availability of offsite resources increases the likelihood of recovery.

3.4.6 Heavy Load Drops

The staff investigated the frequency of a heavy load drop in or near the SFP and the potential
damage to the pool from such a drop. The previous assessment done for resolution of Generic
Issue 82 (in NUREG/CR-4982 (Ref. 4)) only considered the possibility of a heavy load drop on
the pool wall. The assessment conducted for this study identifies other failure modes, such as
the collapse of the pool floor, as also credible for some sites. Details of the heavy load
assessment are given in Appendix 2C. The analysis exclusively considered drops severe
enough to catastrophically damage the SFP so that pool inventory would be lost rapidly and it
would be impossible to refill the pool using onsite or offsite resources. There is no possibility of
mitigating the damage, only preventing it. The staff has not attempted to partition the initiator
into events where there is full rapid draindown and events where there is rapid, but partial
draindown. The staff assumes a catastrophic heavy load drop (creating a large leakage path in
the pool) would lead directly to a zirconium fire. The time from the load drop until a fire varies
depending on fuel age, burn up, and configuration. The dose rates in the pool area before any
zirconium fire are tens of thousands of rem per hour, making any recovery actions (such as
temporary large inventory addition) very difficult.

Based on discussions with staff structural engineers, it is assumed that only spent fuel casks are
heavy enough to catastrophically damage the pool if dropped. The staff assumes a very low
likelihood that other heavy loads will be moved over the SFP and that if one of these lighter
loads over the SFP is dropped, it is unlikely to cause catastrophic damage to the pool.

3-16


CEUS plants. However, HCLPF capacities this high are sufficiently high to achieve seismic risk
estimates less than 3x10* for most CEUS plants based upon the Ref. 8 hazard curves. This
subject is further discussed in Section 4.

In lieu of using a simplified approximate method, Ref. 2 has estimated the seismic risk of
spent fuel pool failure by rigorous convolution of the seismic fragility and seismic hazard
estimates for the 69 CEUS sites for which seismic hazard curves are given in Ref. 8. Ref. 2 has
divided the sites into 26 BWR sites and 43 PWR sites.

For the 26 BWR sites, Ref. 2 used the fragility curve defined in Ref. 5 for Vermont
Yankee with the following properties:

BWR Sites
Median Capacity Cosy = 1.4 PGA
HCLPF Capacity Cucurr = 0.48g PGA (4)

Using the Ref. 8 seismic hazard estimates and the Eqn. (4) fragility, Ref. 2 obtained spent fuel
pool mean annual failure probabilities ranging from 12.0x10° to 0.11x10* and averaging
1.6x10° for the 26 BWR sites. In my judgment, seismic screening criteria (seismic check lists)
can be developed which are sufficiently stringent so as to provide reasonable assurance that the
seismic capacity of spent fuel pools which pass the seismic screening roughly equals or exceeds
that defined by Eqn. (4). With such a fragility estimate, based on the Ref. 8 seismic hazard
estimates, for most CEUS sites, the estimated spent fuel pool seismic-induced failure probability
will be less than 3x10° as further discussed in Section 4.

For the 43 PWR sites, Ref. 2 used the fragility curve defined in Ref. 5 for Robinson with
the following properties:

PWR Sites
Median Capacity Cs = 2.0 PGA
HCLPF Capacity Cucrpr = 0.65g PGA (5)

Using the Ref. 8 seismic hazard estimates and the Eqn. (5) fragility, Ref. 2 obtained spent fuel
pool mean annual failure probabilities ranging from 2.5x10° to 0.03x10° and averaging
0.48x10° for the 43 PWR sites. A fragility curve as high as that defined by Eqn. (5) is necessary
to achieve an estimated spent fuel pool seismic-induced failure probability as low as 1x10° for
nearly all CEUS sites. However, I don't believe realistic seismic screening criteria can be
developed which are sufficiently stringent to provide reasonable assurance that the Eqn. (5)
seismic fragility is achieved. In my judgment, a more rigorous seismic margin evaluation
performed in accordance with the CDFM method described in Refs. 6 or 7 would be required to
justify a HCLPF capacity as high as that defined by Eqn. (5).

A2B-13


Assumptions

For the purpose of the boiloff analysis, the fuel burnup is assumed to be 62.5 GWD/MTU with a
2 year cycle time. The decay heat at this value of burnup is an extrapolation of the decay heat
tables from NUREG/CR-5625 (Ref. 1). The BWR pool is assumed to hold 4,200 9x9 fuel
assemblies. The BWR pool surface area is assumed to be 105.7 square meters. The PWR
pool is assumed to hold 965 17x17 fuel assemblies. The PWR pool surface area is assumed to
be 61.3 square meters. The pools are assumed to have a water depth of 11.54 meters and are
assumed to be at an initial temperature of 30 °C. An estimated water volume fraction of 0.5 of
water in the racks and assemblies is used in the calculations. The specific heat of water is
assumed to be constant at 4,200 J/kg for the heatup calculation. Temperature dependent
properties were used for steel, zircaloy, and UO,. The enthalpy change because of vaporization
used in the boiloff calculation is 2,257 KJ/kg. The results of these calculations are shown in
Table A1-1. The results show that the progression of heatup and boiloff accidents take place on
a very long time scale.

Table A1-1 Heatup and Boiloff Times from Normal Pool Level to Three Feet Above Active Fuel

Decay Time Boiloff Time (hours)
PWR BWR

60 days 100 145

1 year 195 253

2 years 272 337

5 years 400 459

10 years 476 532

1.2 Spent Fuel Heatup Analyses

Once the spent fuel is uncovered (partially or fully), it would begin to heat up. The only
significant heat source initially would be the decay heat. Later, at high cladding temperatures,
additional heat is added by the exothermic oxidation of the zirconium fuel rod cladding. The
staff's review of previous analysis of spent fuel heatup in air concluded that changes to SFP
storage practices indicated a need for significant revisions to previous analysis assumptions.
Accordingly, new analyses were performed for this study to predict fuel rod heatup, in order to
better represent current decommissioning plant operation and SFP storage practices. These
analyses had two basic objectives:

4. Determine the heatup time of fue! cladding from 30 °C to 900 °C ( the temperature at which
the onset of significant fission product release is expected).

2. Determine a generic critical decay time (the time after shutdown that a release of fission
products is no longer possible).

Appendix 1A A1A-2


Part IV. DEPENDENCY

For all tasks, except the first task in the sequence, use the table and formulae below to calculate the
Task Failure Probability With Formal Dependence (P,,,).

If there is a reason why failure on previous tasks should not be considered, explain here:

Dependency Condition Table

Crew Time =: Location Cues Dependen Number of Human Action Failures
(same or : (close in (same or : (additional cy Rule
different) : time or not ornot
close in additional) ~ Not Applicable. Why?.
time
Same Close - complete If this error is the 3rd error in the

sequence, then the dependency is at least
moderate.

If this error is the 4th error in the
sequence, then the dependency is at least
high.

This rule may be ignored only if there is
compelling evidence for less dependence
with the previous tasks. Explain above.

Using P,,,.4= Probability of Task Failure Without Formal Dependence (calculated in Part III, p. 3):
For Complete Dependence the probability of failure is 1.
For High Dependence the probability of failure is (1+ P,,,,)/2
For Moderate Dependence the probability of failure is (1+6 x P,j4)/7

For Low Dependence the probability of failure is (1+19 x P,,,.4)/20



10.

_ U.S. Nuclear Regulatory Commission, “ Operating Experience Feedback

Report—Assessment of Spent Fuel Cooling,” NUREG-1275, Volume 12, February 1997.

Sailor, et.al., “Severe Accidents in Spent Fuel Pools in Support of Generic Issue 82,"
NUREG/CR-4982 (BNL-NUREG-52093), July 1987.

U.S. Nuclear Regulatory Commission, “Control of Heavy Loads at Nuclear Power Plants,
Resolution of Generic Technical Activity A-36,” NUREG-0612, July 1980.

U.S. Department of Energy,“Accident Analysis for Aircraft Crash Into Hazardous Facilities,”
DOE-STD-3014-96, October 1996.

Brookhaven National Laboratory,“Tornado Damage Risk Assessment,” NUREG/CR-2944,
September 1982.

U.S. Nuclear Regulatory Commission,“A Review of the Technical Issues of Air Ingression
During Severe Reactor Accidents,” NUREG/CR-6218, September 1994.

U.S. Nuclear Regulatory Commission,“Code Manual for MACCS2,” NUREG/CR-6613, May
1998.

Electric Power Research Institute,“Seismic Hazard Methodology for the Central and
Eastern United States,” EPRI NP-4726, November 1988.

References for Section 4.0

1.

U.S. Nuclear Regulatory Commission, "Use of Probabilistic Risk Assessment Methods in
Nuclear Regulatory Activities; Final Policy Statement," Federal Register, Vol. 60, No. 158,
August 16, 1995.

U.S. Nuclear Regulatory Commission, “An Approach for Using Probabilistic Risk
Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,”
Regulatory Guide 1.174, July 1998.

U.S. Code of Federal Regulations, “Safety Goals for Operations of Nuclear Power Plants;
Policy Statement,” Part 50, Title 10, “Energy,” August 4, 1986.

Lynnette Hendricks, Nuclear Energy Institute, letter to Office of Nuclear Reactor
Regulation, November 12, 1999.

U.S. Code of Federal Regulations, “Emergency Plans,” Section 47, Part 50, Title 10,
“Energy.”

U.S. Code of Federal Regulations, “Emergency Planning and Preparedness For Production
and Utilization,” Appendix E, Part 50, Title 10, “Energy.”

6-2


2.0 THERMAL-HYDRAULIC ANALYSES

Analyses were performed to evaluate the thermal-hydraulic characteristics of spent fuel stored in
the spent fuel pools (SFPs) of decommissioning plants and determine the time available for
plant operators to take actions to prevent a zirconium fire. These are discussed in Appendix 1A.
The focus was the time available before fuel uncovery and the time available before the
zirconium ignites after fuel uncovery. These times were utilized in performing the risk
assessment discussed in Section 3.

To establish the times available before fuel uncovery, calculations were performed to determine
the time to heat the SFP coolant to a point of boiling and then boil the coolant down to 3 feet
above the top of the fuel. As can be seen in Table 2.1 below, the time available to take actions
before any fuel uncovery is 100 hours or more for an SFP in which pressurized-water reactor
(PWR) fuel has decayed at least 60 days.

Table 2.1 Time to Heatup and Boiloff SFP Inventory Down to 3 Feet Above Top of Fuel

(60 GWD/MTU)
DECAY TIME PWR BWR
60 days 100 hours (>4 days) 145 hours (>6 days)
1 year 195 hours (>8 days) 253 hours (>10 days)
2 years 272 hours (>11 days) 337 hours (>14 days)
5 years 400 hours (>16 days) 459 hours (>19 days)
10 years 476 hours (>19 days) 532 hours (>22 days)

The analyses in Appendix 1A determined that the amount of time available (after complete fuel
uncovery) before a zirconium fire depends on various factors, including decay heat rate, fuel
burnup, fuel storage configuration, building ventilation rates and air flow paths, and fuel cladding
oxidation rates. While the February 2000 study indicated that for the cases analyzed a required
decay time of 5 years would preclude a zirconium fire, the revised analyses show that it is not
feasible, without numerous constraints, to define a generic decay heat level (and therefore
decay time) beyond which a zirconium fire is not physically possible. Heat removal is very
sensitive to these constraints, and two of these constraints, fuel assembly geometry and spent
fuel pool rack configuration, are plant specific. Both are also subject to unpredictable changes
as a result of the severe seismic, cask drop, and possibly other dynamic events which could
rapidly drain the pool. Therefore, since the decay heat source remains nonnegligible for

many years and since configurations that ensure sufficient air flow? for cooling cannot be

2Although a reduced air flow condition could reduce the oxygen levels to a point where a
fire would not be possible, there is sufficient uncertainty in the available data as to when this
level would be reached and if it could be maintained. It is not possible to predict when a

2-1


risk assessment that has a fuel uncovery frequency less than the PPG if they desire an
exemption from EP when their sites are in decommissioning.

Appendix 2B A2B-5


plants, Surry and Peach Bottom,“ for which risk results for both internal and seismic events are
given in NUREG-1150.

Figures 3.7-5 and 3.7-6 show the risk contribution from cask drop events, which are the only
events modeled that are significantly impacted by EP. For the case in which current EP
requirements are retained, the consequences at 1 year and beyond are based on early
evacuation (the lower, solid curve). For the case in which current EP requirements are relaxed,
the consequences within the first 2 years following shutdown are based on late evacuation (the
upper, solid curve), and the consequences at 5 years and beyond are based on early
evacuation, as discussed in Section 3.7.2.

4 The LLNL seismic risk results reported in NUREG-1150 are based on a 1989 version
of the LLNL hazard estimates. An update of these estimates performed in 1993 resulted in a
factor of 10 reduction in the LLNL mean hazard for Peach Bottom and a smaller reduction for
Surry. To provide a more meaningful comparison, the LLNL seismic risk results for Peach
Bottom reported in NUREG-1150 have been reduced by a factor of 10. The results for Surry
and the EPRI seismic risk results are not affected by this adjustment.

3-40


preclude any significant radiation release. In addition, decommissioning activities are
performed according to procedures, which reduces the possibility of operator error. For
example, 10 CFR 50.120 requires training and qualification in nine categories of personnel
involved with spent fuel pool maintenance and support.

Comment #4: During primary system decontamination, can contaminated solution go
“overboard” and into public waters?

Response: Main plant buildings have a drain system in the event of liquid spills to prevent
possible contamination of public waters or land. Additionally, the licensee performs
decommissioning activities using procedures and training, which reduces the likelihood and
consequences of such occurrences.

Comment #5: The NRC should identify the scenario where a steam explosion is possible
because of a severe criticality event and the basis upon which the probability was determined
to be “highly unlikely.”

Response: The staff did not construct a sequence of events that would lead to a steam
explosion. The discussion in the paper (Appendix 3) indicates that the likelihood of any
criticality event is low. The likelinood of a super-prompt critical event is even lower, particularly
in view of the inherent negative feedback provided by fuel temperature increases.

Comment #6: The NRC should identify all radioactivity in the SFP and is capable of being
dispersed in an accident (beyond that on p A3-11 to A3-13).

Response: In Appendix 3, the staff addresses nuclides that contribute to the reactivity of the
spent fuel and does not relate to the generation of the source term. The nuclides listed there
represent well over 90 percent of the reactivity contribution in spent fuel. Therefore, it is not
necessary to expand the list because such an expansion will not significantly alter the predicted
reactivity of the spent fuel in the storage racks. Similarly, nuclides used for the source term,
which is addressed in Appendix 4, represent the dominant contributors to public and worker
dose. The results would not vary significantly if all of the radionuclides were included.

Comment #7: The criticality accident analysis does not consider the risk of a criticality accident
that arises from placement of low-bumup fuel assemblies in a pool where the licensee relies on
bumup credit to prevent criticality.

Response: The scenario suggested by the commenter would only occur as a misloading event
in a pressurized water reactor spent fuel pool. In order to meet General Design Criterion 62,
licensees analyze a misloading event in the spent fuel pool to demonstrate that the fuel
remains subcritical even for the misloading of a fuel assembly of highest reactivity worth (i.e., a
fresh, unirradiated fuel assembly) into a region designed to accommodate lower worth
assemblies (i.e., highly irradiated fuel). Further, the staff demonstrated via analysis (affidavit of
A. Ulses in hearing before the Atomic Safety and Licensing Board, ASLBP No. 99-762-02-LA,
January 4, 2000) that there was sufficient soluble boron in a pressurized water reactor spent
fuel pool such that even an inadvertent misloading of a complete rack of fresh assemblies
would not lead to a criticality event. Therefore, the staff did not give further consideration to this
scenario.

Appendix 6 AG-2


However, given the time scales involved, and the simplicity of the systems, we believe that the
conclusions of this study, namely that, when the NEI commitments are appropriately
implemented the risks are low, are robust.

Certain assumptions may be identified as having the potential for significantly influencing the
results. For example, the calculated time windows associated with the loss of inventory event
tree are sensitive to the assumptions about the leak size. The SPAR HRA method is, however,
not highly sensitive to the time windows assumed, primarily making a distinction between time
windows that represent an inadequate time, barely adequate, nominal, extra time, and
expansive time. The precise definitions of these terms can be found in Reference 9.
Consequently, the assumption of the large leak rate as 60 gpm is not critical. For the loss of
inventory event tree, the assumption that the leak is self-limiting after a drop in level of 15 feet,
may be a more significant assumption that, on a site specific basis may be non-conservative,
and requires validation. The assumption that the preparation time of several days is adequate
to bring offsite sources to bear may be questioned in the case of extreme conditions. However,
the very conservative assumption that this is guaranteed to fail would change the corresponding
event sequences by about an order of magnitude, which would still be a very low risk
contributor.

5.2 Conclusions

The analysis shows that, based on the assumptions made, the frequency of fuel uncovery from
the loss of cooling, loss of inventory, loss of offsite power and fire initiating events is very low.
The assumptions that have been made include that the licensee has adhered to NEI
commitments 2, 4, 5, 8 and 10. In order to take full credit for these commitments, additional
assumptions concerning how these commitments will be implemented have been made. These
include: procedures and/or training are explicit in giving guidance on the capability of the fuel
pool makeup system, and when it becomes essential to supplement with alternate higher
volume sources; procedures and training are sufficiently clear in giving guidance on early
preparation for using the alternate makeup sources; walk-downs are performed on a regular,
(once per shift) basis. The latter is important to compensate for potential failures to the
instrumentation monitoring the status of the pool.

NEI commitment 3, related to establishing communication between onsite and offsite
organizations during severe weather, is also important, though its importance is somewhat
obscured in the analysis by the assumption that there is some degree of dependence between
the decision to implement supplemental make-up to the SFP from onsite sources such as fire
water pumps, and that from offsite sources. However, if no such provision were made, the
availability of offsite resources could become more limiting.

NEI commitments 6, 7 and 9 have been credited with lowering the initiating event frequency for
the loss of inventory events from its historical levels.

This analysis has, demonstrated to the staff that, given an appropriate implementation of the
NEI commitments, the risk is indeed low, and would warrant consideration of granting
exemptions. Without credit for these commitments, the risk will be more than an order of
magnitude higher.

Appendix 2A A2A-67


References:

1. NUREG/CR-4982, Severe Accidents in Spent Fue! Pools in Support of Generic Issue 82,
July 1987.

2. NUREG/CR-6451, A Safety and Regulatory Assessment of Generic BWR and PWR
Permanently Shutdown Nuclear Power Plants, August 1997.

3. NUREG/CR-6613, Code Manual for MACCS2, May 1998.

Appendix 4 A4-17


understatement of the low temperature behavior. Breakaway scaling in an isothermal test may
not translate to similar behavior under transient heatup conditions where initial oxidation occurs
at lower temperatures and may involve steam oxidation. The presence of hydrides in the
cladding may also increase the potential for exfoliation and a breakaway in the oxidation; the
effect of this has, however, been seen more clearly in testing conducted with steam and high
hydrogen concentrations. Also, zirconium hydride will be dissolved at 700 °C and above, thus
its contribution to exfoliation and breakaway will be minimal.

Autoignition is known to occur in zirconium alloys and zirconium hydride, especially when clean
metal or hydride is suddenly exposed to air. The temperature of ignition is highly dependent on
the ratio of surface area to volume and the degree of surface cleanliness. Generally, spent fuel
rod cladding is covered with a relatively thick oxide layer (20-100ym), therefore, unless
ballooning and burst occur in the cladding during heatup, clean high-temperature Zircaloy metal
will not be exposed to air in an SFP accident. However, if there is cladding failure by ballooning
and burst (expected to occur over a temperature range of 700-850 °C), hot oxide-free clean
metal will be abruptly exposed to air. Zirconium hydride is expected to dissolve into the metal
matrix during the slow heatup to these temperatures. At the moment of burst, some clean
surface area of Zr metal will be exposed to air in the location of the rupture. Although data
applicable to this situation is quite limited, considering the relatively small surface-to-volume
ratio of the exposed metal, likelihood of ignition and subsequent propagation of the burning front
of Zr metal is believed to be small (Ref. 8).

In the CODEX tests annular cladded fuel (in a 9 rod bundle) were heated with an inner tungsten
heater rod to examine fuel degradation, with preoxidized cladding, in an air environment.
Zircaloy oxidation kinetics were evaluated as well as the oxidation of the fuel. In the CODEX
AIT-1 test the early phase of the test involved creating a preoxidation using an argon-oxygen
mixture. The intent was to achieve a controlled preoxidation at a temperature of 900-950 °C, but
it was reported (Ref. 3) that preoxidation was started at a slightly higher temperature than
planned. What subsequently occurred was an uncontrollable temperature escalation up to
approximately 2200 °C before it was cooled with cold argon flow. After restabilization of the
rods at 900 °C air injection was started, electrical heatup commenced, and a second
temperature escalation occurred. In the CODEX AIT-2 test, designed to proceed to a more
damaged state, the preoxidation phase was conducted in an argon/steam mixture at 820 °C and
950 °C (a malfunction occurred during the preoxidation phase resulting in the admission of a
small air flow as well). No temperature escalation was seen during the preoxidation phase.
Following the restabilization of the fuel rods, a linear power increase was started and a
temperature excursion subsequently occurred.

In addition to examining relevant test data RES also looked at determining a temperature based
threshold for temperature escalation in an air environment by determining equivalent heat
generation from steam transient tests. In this exercise we posited that at equivalent heat
generation rates, i.e., accommodating different reaction rates and different heats of reaction for
air and steam, we should be able to predict the corresponding temperature for escalation in air
based on temperature escalations seen in severe fuel damage tests conducted in steam. Using
this approach, the heat generation rate was estimated, assuming parabolic kinetics, and the
following equation for a rate constant in air:

Appendix 1B A1B-3


Table 6 Cases with Different Ruthenium-Group Isotopes Included

Case Ruthenium Isotopes Included Mean Consequences (within 100
Release miles)
Fraction
Prompt Societal Cancer
Fatalities | Dose Fatalities
(person-
rem)
Base Case 2x10 Co-58,Co-60,Ru-103,Ru-106 1.01 4.54x10° 2,320
11 1 Co-58,Co-60,Ru-103,Ru-106 95.3 9.53x10° 9,150
dla 1 Ru-103,Ru-106 94.4 9.51x10® 9,120
11b 41 Ru-106 94.3 9.51x10° 9,120
tic 1 Ru-103 1.02 4.54x10° 2,320

The amounts of the dominant cesium isotope, Cs-137, and the dominant ruthenium isotope, Ru-106, ina
spent fuel pool at one year after final shutdown are about the same. After one year, the inventories of
Cs-137 and Ru-106 are 8.38x10"” Bq and 5.77x10" Bq, respectively. This would suggest a modest
increase in the individual risk of a prompt fatality ruthenium is included in the consequence calculation.
However, Table 4 shows large increases in the individual risk of a prompt fatality. A comparison of the
dose conversion factors for Cs-137 and Ru-106 is given in Table 7. These dose conversion factors were
taken from the MACCS input file DOSDATA.INP. An examination of these dose conversion factors
indicates that the large Ru-106 inhalation dose conversion factor in MACCS used to calculate acute
doses is partly responsible for the increase in individual risk of a prompt fatality beyond what would be
expected as a result of the additional amount of Ru-106.

Table 7 Dose Conversion Factors for Ru-106 and Cs-137

organ cloud- ground- inhalation/ | inhalation/ | ingestion
shine shine acute chronic (Sv/Bq)

(Svsec! | (Svsec/ | (Sw/Bq) (Sw/Ba)
Bq m°) Bq m’)

Ru-106 lungs 7.99E-15 1.58E-16 | 2.09E-08 1.04E-06 1.48E-09

red marrow 8.05E-15 1.61E-16 | 8.74E-11 1.77E-09 1.48E-09

Cs-137 lungs 2.88E-14 4.35E-16 | 8.29E-10 8.80E-09 1.27E-08

red marrow 2.22E-14 4.41E-16 | 5.63E-10 8.30E-09 1.32E-08

Ratio of Ru-106 lungs 4 4 25 118 A
to Cs-137
red marrow A 4 2 2 1
Fuel Fines:

The staff performed MACCS calculations with different fuel fines release fractions to assess the
sensitivity of the consequences. The results of these calculations are shown in Table 8. Case 11, which
used a ruthenium release fraction of one, is the shown in the second row of Table 8 and was the starting
point for these calculations. Then, Case 96 was run with the large fuel fines release fraction of .01. Asa
result of increasing the fuel fines release fraction from 1x10* to .01, a small increase in the offsite

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A-12



within a cover plate tend to prevent any appreciable fragmentation and movement of the
enclosed absorber material. The total loss of the welded cover plate is not considered
feasible.

Boraflex has been found to degrade in spent fuel pools because of gamma radiation and
exposure to the wet pool environment. For this reason, the NRC issued Generic

Letter 96-04 to all holders of operating licenses, on Boraflex degradation in spent fuel
storage racks. Each addressee that uses Boraflex was requested to assess the capability
of the Boraflex to maintain a 5 percent subcriticality margin and to submit to the NRC
proposed actions to monitor the margin or confirm that this 5 percent margin can be
maintained for the lifetime of the storage racks. Many licensees subsequently replaced the
Boraflex racks in their pools or reanalyzed the criticality aspects of their pools, assuming
no reactivity credit for Boraflex.

Other potential criticality events, such as loose debris of pellets or the impact of water or
firefighting foam (adding neutron moderation) during personnel actions in response to
accidents, were discounted because of the basic physics and neutronic properties of the racks
and fuel, which would preclude criticality conditions being reached with any creditable
likelihood.

For example, without moderation, fuel at current enrichment limits (no greater than 5 wt percent
U-235) cannot achieve criticality, no matter what the configuration. If it is assumed that the pool
water is lost, a reflooding of the storage racks with unborated water or fire-fighting foam may
occur because of personnel actions. However, both PWR and BWR storage racks are
designed to remain subcritical if moderated by unborated water in their normal configuration.
The phenomenon of a peak in reactivity because of low-density (optimum) moderation (fire-
fighting foam) is not of concern in spent fuel pools since the presence of relatively weak
absorber materials, such as stainless steel plates or angle brackets, is sufficient to preclude
neutronic coupling between assemblies. Therefore, personnel actions to refill a drained spent
fuel pool containing undeformed fuel assemblies would not create the potential for a criticality.
Thus, the only potential scenarios described above in 1 and 2 involve crushing of fuel
assemblies in low-density racks or degradation of Boraflex over long periods in time.

To gain qualitative insights on the criticality events that are credible, the staff considered the
sequences of events that must occur. For scenario 1 above, this would require a heavy load
drop into a low-density racked BWR pool compressing assemblies. From Appendix 2C on
heavy load drops, the likelihood of a heavy load drop from a single failure-proof crane is
approximately 2E-6 per year, assuming 100 cask movements per year at the decommissioning
facility. From the load path analysis done for that appendix, it was estimated that the load could
be over or near the pool between 5 percent and 25 percent of the movement path length,
dependent on plant-specific layout specifics. The additional frequency reduction in the
appendix, to account for the fraction of time that the heavy load is lifted high enough to damage
the pool liner, is not applicable here because the fuel assemblies could be crushed without the
same impact velocity being required as for the pool floor or wall. Therefore, if we assume

10 percent load path vulnerability, we observe a potential initiating frequency for crushing of
approximately 1.2E-6 per year (based upon 100 lifts per year). Criticality calculations in this
appendix show that even if the low-density BWR assemblies were crushed by a transfer cask, it
is "highly unlikely" that a configuration would be reached that would result in a severe reactivity

Appendix 3 A3-2


W.C. Jordan, et. al., "Validation of KENOV.a Comparison with Critical Experiments,”

ORNL/CSD/TM-238, Oak Ridge. National Laboratory, Oak Ridge National Laboratory,
1986.

“Licensing Report for Expanding Storage Capacity in Harris Spent Fuel Pools C and D,"
HI-971760, Holtec International, May 26, 1998, (Holtec International Proprietary)

Appendix 3 A3-9


5.0 SUMMARY OF RESULTS

The results of this analysis provide insight into the risks associated with storage of spent nuclear
fuel in fuel pools at decommissioned nuclear power plants. The five accident initiators that were
analyzed consist of: 1) internal fires, 2) loss of cooling, 3) loss of inventory, 4) plant/grid
centered losses of offsite power, and 5) severe weather induced losses of offsite power. The
total frequency for the endstate is estimated to be 1.8E-7/year. Table 5.1 summarizes the fuel
uncovery frequency for each initiator.

This frequency is to be compared with the pool performance guideline (PPG). This guideline
has been established by analogy with the acceptance guidelines in RG. 1.174. In RG 1.174 it
was determined that the mean value of the distribution characterizing uncertainty is the
appropriate value to compare the guideline. However, it was determined that it is also
necessary to investigate whether there are modeling uncertainties that could affect the decision
made with respect to whether the guidelines have been met. This is the approach that has been
followed here.

5.1 Characterization of Uncertainty

The frequencies are point estimates, based on the use of point estimates for the input
parameters. The input parameter values were taken from a variety of sources, and in many
cases were presented as point estimates with no characterization of uncertainty. In some
cases, such as the initiating event frequencies derived from NUREG/CR 5496, and the HEPs
derived from THERP, an uncertainty characterization was given, and the point estimates chosen
corresponded to the mean values of the distributions characterizing uncertainty. For all other
parameters, it was assumed that the values would be the mean values of distributions
characterizing the uncertainty of the parameter value. In the case of SPAR HEPs, the authors
of the SPAR HRA approach consider their estimates as mean values based on the fact that the
numbers were established on the basis of considering several different sources, most of which
specified mean values. Consequently, the results of this analysis are interpreted as being mean
values. A propagation of parameter uncertainty through the model was not performed, nor was
it considered necessary. With the exception of the SFP cooling system itself, the systems relied
on are single train systems. The dominant failure contributions for the SFP cooling system are
assumed to be common cause failures. Thus there are no dominant cutsets in the solutions that
involved multiple repetitions of the same parameter, and under these conditions, use of mean
values as input parameters produces a very close approximation to mean values of sequence
frequencies. Since typical uncertainty characterization for the input parameters is a lognormal
distribution with error factors of 3 or 10, the 95" percentile of the output distribution will be no
more than a factor of three higher than the mean value. This is not significant to change the
conclusion of the analysis.

The numerical results are a function of the assumptions made and in particular, the model used
to evaluate the human error probabilities. The staff believes the models used are appropriate
for the purpose of this analysis, and in particular are capable of incorporating the relevant
performance shaping factors to demonstrate that low levels of risk are achievable, given an
appropriate level of attention to managing the facility with a view to ensuring the health and
safety of the public. Alternate HRA models could result in frequencies that are different.

Appendix 2A A2A-66


Table A2d-3  DWTF Aircraft Crash Hit Frequency (per year)

Period Air Carriers Air Taxes General Aviation Military Aviation Total

1995, 1.72x107 2.47x10° 2.45x10% 5.03x107 2.76x10°
1993-1995 1.60x107 2.64x10° 2.82x10% 6.47x107 3.16x10°
1991-1995 1.57x107 2.58x10° 2.89x19 7.23x107 3.23x10
1986-1995 4.52x107 2.41x10° 2.89x10° 8.96x107 3.23x105

Note (1): Various periods were studied to assess variations in air field operations.

Table A2d-4 Probability of Penetration as a Function of Location and Concrete Thickness

Probability of penetration
Thickness of reinforced concrete

Plant location Aircraft type 1 foot 1.5 feet 2 feet 6 feet
s 5 miles ‘Small < 12,000 Ibs 0.003 0 Oo 0
from airport

Large > 12,000 Ibs 0.96 0.52 0.28 0
> Smiles Small < 12,000 Ibs 0.28 0.06 0.01 0
from airport

Large > 12,000 Ibs 1.0 1.0 0.83 0.32

Appendix 2D A2D-6



APPENDIX 5
NOVEMBER 12, 1999 NUCLEAR ENERGY INSTITUTE COMMITMENT LETTER

Appendix 5 A5-1


Table 4.1-2 Staff Decommissioning Assumptions (SDAs)

SDA No. Staff Assumptions

1 Licensee’s SFP cooling design will be at least as capable as that assumed in
the risk assessment, including instrumentation. Licensees will have at least
one motor-driven and one diesel-driven fire pump capable of delivering
inventory to the SFP.

2 Walk-downs of SFP systems will be performed at least once per shift by the
operators. Procedures will be developed for and employed by the operators
to provide guidance on the capability and availability of onsite and offsite
inventory makeup sources and time available to initiate these sources for
various loss of cooling or inventory events.

3 Control room instrumentation that monitors SFP temperature and water level
will directly measure the parameters involved. Level instrumentation will
provide alarms at levels associated with calling in offsite resources and with
declaring a general emergency.

4 Licensee determines that there are no drain paths in the SFP that could lower
the pool level (by draining, suction, or pumping) more than 15 feet below the
normal pool operating level and that licensee must initiate recovery using
offsite sources.

5 Load Drop consequence analyses will be performed for facilities with non-
single failure-proof systems. The analyses and any mitigative actions
necessary to preclude catastrophic damage to the SFP that would lead to a
rapid pool draining would be sufficient to demonstrate that there is high
confidence in the facilities ability to withstand a heavy load drop.

6 ach decommissioning plant will successfully complete the seismic checklist
provided in Appendix2B to this study. If the checklist cannot be successfully
completed, the decommissioning plant will perform a plant specific seismic
risk assessment of the SFP and demonstrate that SFP seismically induced
structural failure and rapid loss of inventory is less than the generic bounding
estimates provided in this study (<1x10° per year including non-seismic
events).

7 Licensees will maintain a program to provide surveillance and monitoring of
Boraflex in high-density spent fuel racks until such time as spent fuel is no
longer stored in these high-density racks.

4-12



The site-specific area is shown in Figure A2d-1 and is further defined as:

Pept = Ag t+ As
where:
: Equation A2d-2
2:L-W-WS
Ay = (WS +R)-(H-cota)+ =E-ZENS bw
A s7 (WS +R)-S
and where:
Acs = total effective target area H= height of facility
A, = effective fly-in area L= length of facility
A, = effective skid area We width of facility
WS= wing span S= aircraft skid distance
cot0= mean of cotangent of aircraft R= length of facility diagonal
impact angle

Alternatively, a point target area model was defined as the area (length times width) of the
facility in question, which does not take into account the size of the aircraft.

Table A2d-1 summarizes the generic aircraft data and crash frequency values for five aircraft
types (from Tables B-14 through B-18 of DOE-STD-3014-96). The data given in Table A2d-1
were used to determine the frequency of aircraft hits per year for various building sizes (length,
width, and height) for the minimum, average, and maximum crash rates. The resulting
frequencies are given in Table A2d-2. The product N,.*P).“f(x,y) for Equation A2d-1 was taken
from the crashes per mi*/yr and A, was obtained from Equation A2d-2 for aircraft characteristics.
Two sets of data were generated: one included the wing and skid lengths, using the effective
aircraft target area model, and the other considered only the area (length times width) of the
site, using the point target area model.

The results from the DOE effective aircraft target area model, using the generic data in

Table A2d-1, were compared to the results of two evaluations reported in Reference 2. The first
evaluation of aircraft crash hits was summarized by C.T. Kimura et al. in Reference 3. The
DWTF Building 696 was assessed in the Kimura report. It was a 1-story 254-feet-long 80-feet-
wide, 39-feet-high structure. The results of Kimura’s study are given in Table A2d-3.

Applying the DOE generic data to the DWTF resulted in a frequency range of 6.5x1 0° hits per
year to 6.6x10° hits per year, with an average value of 4.4x10° per year, for the effective aircraft
target area model. For the point target area model, the range was 4.4x10" to 2.2x10° per year,
with an average value of 1.5x107 per year.

Appendix 2D A2D-2


For all three approaches the Screening Level 2 HCLPF capacity defined by Eqn. (8) was used. In
addition, for both the Ref. 10 and rigorous convolution approaches, a fragility logarithmic

standard deviation B of 0.4 was used.

From Table 2, it can be seen that the Ref. 1 method (Eqn. (3) ) underestimates the seismic
tisk by factors of 2.3 and 3.5 for Vermont Yankee and Robinson, respectively. The simplified
approach recommended in Ref. 10 and described herein overestimates the seismic risk by 20%
and 5% respectively for these two cases. These results are consistent with the results I have
obtained for many other cases.

4.2 Estimated Seismic Risk of Spent Fuel Pools Screened at Screening Level 2 Using
Mean LL93 Hazard Estimates from Ref. 8 and 9

Using the Ref. 10 simplified approach described in the previous subsection, I have
estimated the spent fuel pool seismic risk of failure corresponding to Screening Level 2 for all 69
CEUS sites with LLNL93 seismic hazard estimates defined in Refs. 8 and 9. These sites are
defined in terms of an NRC site number code (OCSP_) used in Ref. 9. For each site, I assumed
that the HCLPF capacity Cycrpp was defined by Eqn. (8). A total of 35 of the 69 sites had
estimated seismic risks of spent fuel pool failure associated with Screening Level 2 of greater
than 1x10-6. The estimated seismic risk of 26 of these sites exceeded 1.25x10*. These 26 sites
with their estimated seismic risk corresponding to Screening Level 2 are listed in Table 3. As can
be seen in Table 3, only 8 of the 69 sites had estimated seismic risks of spent fuel pool failure
exceeding 3x10°. One of these sites is Shoreham at which no fuel exists.

It should be noted that the seismic risks of spent fuel pool failure tabulated in Table 3 are
based on the assumption that the HCLPF capacity of the spent fuel pool exactly equals the
Screening Level 2 HCLPF capacity of 1.2g PSA (equivalent to 0.5g PGA). In actuality, spent
fuel pools which pass the appropriately defined screening criteria are likely to have capacities
higher than the screening level capacity. Therefore these are upper bound seismic risk estimates
for spent fuel pools that pass the to-be established screening criteria. Furthermore, the simplified
approach used to estimate the seismic risks in Table 3 overestimates these risks by 0% to 25%.

4.3 Estimated Seismic Risk of Spent Fuel Pools Screened at Screening Level 2 Using
Mean EPRI89 Hazard Estimates

Following the exact same Ref. 10 simplified approach which I followed for the LLNL93
hazard estimates, Ref. 11 provides the corresponding seismic risk of spent fuel pool failure
estimates based upon EPRI89 hazard estimates for 60 of the 69 CEUS sites. Table 3 shows the
corresponding seismic risk computed in Ref. 11 for the EPRI89 hazard estimates.

From Table 3, it can be seen that the EPRI89 hazard estimates produce generally much
lower seismic risk estimates corresponding to Screening Level 2 than do the LLNL93 hazard
estimates. Based on the EPRI89 hazard estimates, only one site has a seismic risk exceeding
1x10. Only three other sites have seismic risks exceeding 0.5x10*. Table 3 includes all sites

A2B-18


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3-22


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10.

11.

References

Preliminary Draft Technical Study of Spent Fuel Pool Accidents for
Decommissioning Plants, Nuclear Regulatory Commission, June 16, 1999

Draft EPRI Technical Report: Evaluation of Spent Fuel Pool Seismic Failure
Frequency in Support of Risk Informed Decommissioning Energy Planning, Duke
Engineering and Services

A Review of Draft NRC Staff Report: Draft Technical Study of Spent Fuel Pool
Accidents for Decommissioning Plants, NEI, August 27, 1999

Seismic Screening Criteria for Assessing Potential Fuel Pool Vulnerabilities at
Decommissioning Plants, NEI, August 18, 1999

Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two
Representative Nuclear Power Plants, NUREG/CR-5176, Prepared for Nuclear
Regulatory Commission, January 1989

An Approach to the Quantification of Seismic Margins in Nuclear Power Plants,
NUREG/CR-4334, Prepared for Nuclear Regulatory Commission, August 1985

A Methodology for Assessment of Nuclear Power Plant Seismic Margin
(Revision 1), (EPRI NP-6041-SL), August 1991

Revised Livermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites
East of the Rocky Mountains, NUREG-1488, Nuclear Regulatory Commission,
October 1993

Extension to Longer Return Periods of LLNL Spectral Acceleration Seismic Hazard
Curves for 69 Sites, provided by Engineering Research Applications Branch,
Nuclear Regulatory Commission, September, 1999

Kennedy, R-P., Overview of Methods for Seismic PRA and Margin Assessments
Including Recent Innovations, CSNI Seismic Risk Workshop, Tokyo, Japan,
August 1999

Personal Communication from Tom O'Hara, Duke Engineering and Services to
Robert Kennedy, October 19, 1999

A2B-20


Consequence estimates are also included on Figures 3.7-1 and 3.7-2 for the two operating
reactors for which risk results for both internal and seismic events are available in
NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,”
and the supporting NUREG/CR-4551 reports, “Evaluation of Severe Accident Risks: Surry Unit
1" and “Evaluation of Severe Accident Risks: Peach Bottom, Unit 2.” The values shown are for
the reactor accident source terms that produced the greatest number of early fatalities

(Figure 3.7-1) or the greatest societal dose and latent cancer fatalities (Figure 3.7-2). Results
are displayed separately for internally and seismically initiated accidents and indicate that for
these plants, reactor accident consequences for seismically initiated events are substantially
higher than those for internally initiated events. Although the consequences for the high
ruthenium source term diminish more quickly than for the low ruthenium source term, these
curves do not converge because of the long half-lives of the fuel fines in the high ruthenium
source term.

An examination of Figure 3.7-1 indicates the following:

+ Early fatality consequences for spent fuel poo! accidents can be as large as for a severe
reactor accident even if the fuel has decayed several years. This is attributable to the
significant health effect of ruthenium, and the ruthenium-106 half-life of about 1 year. There
is also an important but lesser contribution from cesium.

+ Alarge ruthenium release fraction is important to consequences, but not more important
than the consequences of a reactor accident large early release.

+ The effect of early evacuation (if possible) is to offset the effect of a large ruthenium release
fraction. This effect is comparable to that for reactor accidents.

+ For the low ruthenium source term, no early fatality is expected after 1 year decay even
with late evacuation.

For the longer term consequences Figure 3.7-2 indicates:

+ Long-term consequences remain significant as long as a fire is possible. These
consequences are due primarily to the effect of cesium-137, which remains abundant even
in significantly older fuel because of its long (30-year) half-life. Ruthenium and evacuation
have notable long-term consequences but do not change the conclusion.

3.7.2 Risk Modeling for SFP Accidents

The quantitative assessment of risk involves combining the estimated frequencies of severe
accident sequences with their corresponding offsite consequences. In this section, severe
accident consequences reported in Tables 3.7-1 and 3.7-2 are assigned to each of the major
types of events that lead to uncovery of the spent fuel, and then combined with the respective
event frequencies to provide a scoping estimate of SFP risks.

The SFP accidents discussed in Section 3 can be broadly classified as either boildown or rapid
draindown sequences. Rapid draindown sequences are further divided into seismically- and

3-34


4.0 MODEL DEVELOPMENT

This section describes the risk models that were developed to assess the likelihood of fuel
uncovery from SFP loss of cooling events, fire events, loss of offsite power, and loss of
inventory events.

4.1 Loss of Cooling Event Tree

This event tree (Figure 4.1) models generic loss of cooling events (i.e., those not related to other
causes such as fire or loss of power, which are modeled in later sections). The top events and
the supporting functional fault trees are discussed in the following sections.

4.1.1 Initiating Event LOC — Loss of Cooling
4.1.1.1 Event Description

This initiating event includes conditions arising from loss of coolant system flow because of the
failure of the operating pumps or valves, from piping failures, from an ineffective heat sink
(e.g., loss of heat exchangers), or from a local loss of power (e.g., failure of electrical
connections).

4.1.1.2 Quantification

This initiating event is modeled by a single basic event, IE-LOC. An initiation frequency of
3.0E-3/yr is taken from NUREG-1275 Volume 12 (Ref. 14). This represents the frequency of
loss of cooling events in which temperatures rise more than 20 °F.

4.1.2 Top Event CRA - Control Room Alarms
4.1.2.1 Event Description and Timing

This event represents a failure to respond to conditions in the pool that are sufficient to trigger
an alarm. Failure could be because of operator error (failure to respond), or loss of indication
because of equipment faults. Success for this event is defined as the operator recognizing the
alarm and understanding the need to investigate its cause. This event is quantified by fault tree
LOC-CRA and includes hardware and human failures basic events that represent failure of
control room instrumentation to alarm given that SFP cooling has been lost, and the operators
fail to respond to the alarm, respectively.

4.1.2.2 Relevant Assumptions

° Within 8 to 12 hours of the loss of cooling, one or more alarms or indications will reflect an
out-of-tolerance condition to the operators in the contro! room (there may be level
indication available locally or remotely, but any change in level is not likely to be significant
until later in the sequence of events).

. The SFP has at least one water temperature measuring device, with an alarm and a

Appendix 2A A2A-13


For the case in which current EP requirements are retained, it was assumed that cask drop
accidents occurring 1 or more years following shutdown would afford sufficient time to
implement protective measures before fission products were released. This is consistent with
the evacuation time estimates in the NUREG-1150 study for Surry, which assumed a 1.5 hour
delay time and a 4 mile per hour evacuation speed. Thus the consequences at less than 1 year
following shutdown are based on late evacuation, and the consequences at 1 year and beyond
are based on early evacuation when full EP requirements are retained.

Relaxations in EP requirements are expected to result in additional delays in initiation and
implementation of protective measures relative to the case in which current EP requirements are
retained. If offsite preplanning requirements were to be relaxed, as many as 10 to 15 hours may
be required at some sites to initiate an evacuation. Based on either air-cooled or adiabatic
heatup rates for the reference pool, the minimum time to fission product release following a load
drop that catastrophically damages the pool is about 8-9 hours for PWR pools and about

15 hours for BWR pools 2 years following shutdown (see Appendix 1A). These release times
increase significantly by 5 years following shutdown (i.e., greater than 24 hours even with
adiabatic heatup rates). For the case in which current EP requirements are relaxed, the
consequences within the first 2 years following shutdown are based on late evacuation, and the
consequences at 5 years and after are based on the early evacuation results reported in

Tables 3.7-1 and 3.7-2.

3.7.3. Risk Results

The frequency and consequences for each SFP accident were combined to provide a scoping
estimate of the risk of SFP accidents. The frequency of each event was based on the estimated
value at 1 year following shutdown as described above, and was assumed to remain constant
over time. In reality, the frequency would vary with time, and could be higher or lower than the
1-year estimate, as a result of plant configuration changes described in Section 3.1 (e.g.,
replacement of operating plant pool cooling and makeup systems with skid-mounted systems)
and reductions in decay heat levels (which would impact human reliability estimates). However,
as described in Section 3.4.7, these impacts are not expected to change the insights from the
tisk assessment for decay times greater than 60 days.

Figures 3.7-3 and 3.7-4 show the total early fatality risk and societal risk as a function of time
after final shutdown. Companion curves are provided based on both the LLNL and the EPRI
seismic hazard studies since both studies are considered equally valid. The SFP risk results are
shown in these figures for both the high ruthenium source term and a fuel burnup of

60 GWD/MTU. Also shown are the corresponding mean risk measures for two operating

3-39


period of an earthquake ground motion that would damage a decommissioning plant's SFP
cooling system equipment (assuming it had at least minimal anchoring) is about once in

4,000 years. The staff quantified a human error probability of 4x10* that represents the failure
of the fuel handlers to obtain offsite resources. The event was quantified using the SPAR HRA
technique. The performance shaping factors chosen were as follows: expansive time (> 50
times the required time), high stress, complex task because of the earthquake and its non-
routine nature, quality procedures, poor ergonomics because of the earthquake, and finally a
crew who had executed these tasks before, conversant with the procedures and one another.
In combination we now estimate the risk from support failure because of seismic events to be on
the order of 4x10® per year. The frequency of fuel uncovery from support system failure
because of seismic events is bounded by catastrophic failure of the SFP because of a seismic
event.

5. HAZARD ESTIMATE AND FRAGILITY UNCERTAINTIES

The staff recognizes there are considerable uncertainties in both the seismic hazard estimates
for nuclear power plant sites and for the fragility estimates of SFPs. The staffs evaluation used
both LLNL and EPRI hazard estimates (frequency of the ground motion occurring, at a certain
level) since the NRC has stated that both the EPRI and LLNL hazard estimates are reasonable
and valid. For eastern U.S. sites, the hazard estimates (particularly LLNL) are relatively flat as
the return period and peak spectral acceleration increase. At the return frequency (ie.
frequency of an earthquake at or exceeding a specified ground motion level) of safe shutdown
earthquakes (SSEs), the LLNL and EPRI estimates are in reasonable agreement. However, as
ground motion levels increase, there is little or no conclusive data, and the ground motion
experts diverge on how to assign return periods to extreme seismic events. The tails of the
hazard curve distributions drive the results (i.e. the mean) as would be expected of a distribution
that is particularly flat (e.g., one that has large modeling uncertainties).

6. CONCLUSION

The staff recommends that those plants that are west of the Rocky Mountains and have short
return periods for the occurrence of ground motions greater than 1.2 g PSA in their SFP should
be required to conduct plant-specific seismic analysis beyond the confirmation of the checklist if
they desire to obtain exemptions (or take advantage of rule making) from EP, indemnification, or
security at decommissioning sites. In addition, because the LLNL hazard estimates indicate that
the H. B. Robinson site exceeds the PPG (1x10° per year) at 1.2g PSA, the utility should
perform a plant-specific seismic analysis too.

To summarize the staff recommendations to provide reasonable confidence that there are no
seismic vulnerabilities at decommissioning plant SFPs, (1) all sites must conduct an assessment
of the SFP structures using the revised seismic check list in order to identify any structural
degradation, potential for seismic interaction from superstructures and overhead cranes, and to
verify that they have a seismic HCLPF value of 1.2 g PSA or higher, (2) those sites that do not
pass the seismic check list may either undertake appropriate remedial action or conduct a site-
specific seismic risk assessment of their decommissioning risk, and (3) sites such as H. B.
Robinson, WNP2, Diablo Canyon, and San Onofre should have to pass the seismic check list to
identify any structural degradation or other anomalies and then conduct a site-specific seismic

Appendix 2B A2B-4


maintenance and switching), localized weather-induced faults (e.g., lightning), or combinations
of these. Grid-related events are those in which problems in the offsite power grid cause the
loss of offsite power.

4.3.1.2 Quantification

For plant-centered LOSP events, NUREG/CR-5496 (Ref. 18) estimates a frequency of
.04/critical year for plant centered loss of offsite power for an operating plant, and .18/unit
shutdown year for a shutdown plant. For grid-related LOSP events, a frequency of 4E-3/site-yr
was estimated. The frequency of grid-related losses is assumed to be directly applicable.
However, neither of the plant centered frequencies is directly applicable. At a decommissioning
plant there will no longer be the necessity to have the multiplicity of incoming lines typical of
operating plants, which could increase the frequency of loss of offsite power from mechanical
failures. On the other hand, the plant will be a normally operating facility, and it would be
expected that there will be less activity and operations in the switchyard than would be expected
at a shutdown plant, which would decrease the frequency of loss from human error, the
dominant cause of losses for shutdown plants. For purposes of this analysis, the LOSP initiating
event frequency of 0.08/yr, assumed in INEL-96/0334 (Ref. 15), is assumed for the combined
losses from plant-centered and grid-related events.

4.3.2 Top Event OPR - Offsite Power Recovery
4.3.2.1 Event Description and Timing

The fault tree for this top event (LP1-OPR) is a single basic event that represents the
non-recovery probability of offsite power.

NUREG-1032 (Ref. 19) classified LOSP events into plant-centered, grid-related, and
severe-weather-related categories, because these categories involved different mechanisms
and also seemed to have different recovery times. Similarly, NUREG/CR-5496 (Ref. 18) divides
LOSP events into three categories and estimates different values of non-recovery as functions
of time.

4.3.2.2 Relevant Assumptions

. Trained electricians may not be present at the site for quick recovery from plant-centered
events.

Appendix 2A A2A-34


the Commission’s QHOs continue to be met for SFP fires even if the ruthenium release fraction
is substantially increased. Given these observations, there does not appear to be sufficient
justification to revise the proposed pool performance guideline of 1x10° per year which was
developed from the RG 1.174 LERF considerations.

Although the above comparisons focus on the Surry site, the results are expected to be
generally applicable to other sites as well. The QHOs represent risk to the average individual
within 1 mile and 10 miles of the plant, and should be relatively insensitive to the site-specific
population.

It should also be acknowledged that long term health impacts are sensitive to public policy
decisions such as land interdiction criteria for returning populations. The long term protective
assumption used in both the NUREG-1150 and SFP studies was to interdict land which could
give a projected dose to an individual via the groundshine and resuspension inhalation
pathways of more than 4 rem in 5 years (2 rem in the first year and 0.5 rem per year for the
next 4 years, for an average of 800 mrem per year). Comparisons of consequence results at
various distances for each of the NUREG-1150 reference plants are provided in NUREG/CR-
6349, and clearly show that the increase in population dose with distance is because of a large
number of people receiving very smail doses, below the assumed long-term interdiction limit of
4 rem in 5 years, since the offsite consequences because of land condemnation, etc., remain
essentially the same over the range of distances. The effect of varying long-term interdiction
dose limits on population doses, latent fatalities, and offsite costs was estimated in
NUREG/CR-6349 by recalculating the consequences for each of the NUREG-1150 plants for
various lower limits. The results show that as the interdiction limit is reduced, the population
dose and latent cancers decrease and the offsite costs progressively increase. For a reduction
in the interdiction limit from 800 mrem per year to 300 mrem per year the risk measures
decreased by typically 20 to 30 percent, and offsite costs increased by about a factor of two.
Thus, changes in risk results on this order can be expected as a result of public policy
decisions.

3. REFERENCES

1. Safety Goals for the Operations of Nuclear Power; Policy Statement, 51_Federal Register
28044, August 4, 1986.

2. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.

3. U.S. Nuclear Regulatory Commission, Evaluation of Severe Accident Risks: Surry Unit 1,
NUREG/CR-4551, Vol. 3, Rev. 1, Part 1, Sandia National Laboratory, October 1990.

4. U.S. Nuclear Regulatory Commission, MACCS Version 1.5.11.1: A Maintenance Release
of the Code, NUREG/CR-6059, Sandia National Laboratory, October 1993.

Appendix 4C A4C-4


non-seismically-initiated events. In assigning consequences to each of these events, the staff
considered whether protective measures to evacuate the population around the site could be
effectively implemented before fission product release. This included consideration of the
effectiveness of offsite notification, the delay between event initiation and fission product release
(dependent on time after shutdown), the time required to initiate and complete an evacuation,
and the impact that a relaxation in current emergency planning requirements might have on
these factors. As a result of this assessment, consequences were assigned based on either the
early evacuation case or late evacuation case.

The frequency and consequence modeling is briefly described below for each type of SFP
accident. The resulting risk estimates for each sequence (in terms of early fatalities and societal
dose per year) are presented in Figures 3.7-3 through 3.7-6 and discussed in Section 3.7.3.

Boil Down Sequences

Boil down sequences (including loss of inventory events) and their associated frequencies are
listed in Table 3.7-3. These sequences involve heatup of the pool to boiling followed by gradual
reduction in pool level until the spent fuel is eventually uncovered. This process would take over
100 hours at 60 days, and substantially longer at later times as shown in Table 2.1. The long
delay provides sufficient time for licensee staff to effectively intervene in the large majority of
these events, and results in very low frequencies of fuel uncovery. For those events that
proceed to fuel uncovery, fuel heatup will continue until either steady-state conditions are
achieved or cladding oxidation occurs. All boil down sequences that uncover spent fuel were
assumed to result in an SFP fire. Loss of inventory events are classified as boil down events
since the time to uncover the fuel will be in excess of 24 hours (as described in Section 4.5.4.1
of Appendix 2A) and will provide ample time for licensee to take corrective measures.

Table 3.7-3 Frequency of Boil Down Events Leading to Spent Fuel Uncovery (for times
greater than 60 days after shutdown)

Initiating Event Frequency (per year)
Loss of offsite power—severe weather 1.1x107
Loss of offsite power—plant-centered and grid-related events 2.9x10°
Internal fire 2.3x10*
Loss of pool cooling 1.4x10%
Loss of coolant inventory 3.0x10°
Total : ° 1.8x107

The failure paths leading to a zirconium fire involve failure to acquire offsite resources to
makeup pool inventory, despite the large amount of time available for recovery in the boildown
event. For sequences involving loss of offsite power due to severe weather, the weather is
assumed to drain regional resources or limit access to the facility. The staff reasoned that if it

3-35


Commission which extends the PSA seismic hazard estimates also down to 10°. Ref. 9 is
attached herein as Appendix A.

In order to achieve a seismic induced annual failure probability P, in the low 10° range
for nearly all of the CEUS spent fuel pools with the Ref. 8 hazard estimates, it is necessary to
apply the Level 2 screening criteria of Refs. 6 or 7, i-e., screen at a HCLPF seismic capacity of
1.2g PSA (equivalent to 0.5g PGA). The seismic screening criteria presented in Ref. 4 is
properly based upon screening to Level 2. Furthermore, Ref. 4 appropriately summarizes the
guidance presented in Ref. 7 for screening to Level 2. In general, I support the screening criteria
defined in Ref. 4. However, I do have three concerns which are discussed in the following
subsections.

3.1 Out-of-Plane Flexural and Shear Failure Modes for Spent Fuel Pool
Concrete Walls and Floor

The screening criteria for concrete walls and floor diaphrams were developed to
provide seismic margin HCLPF capacities based upon in-plane flexural and shear failures of
these walls and diaphrams. For typical auxiliary buildings, reactor buildings, diesel
generator buildings, etc., it is these in-plane failure modes which are of concern. For normal
building situations, seismic loads are applied predominately in the plane of the wall or floor
diaphram. Out-of-plane flexure and shear are not of significant concern. As one the primary
authors of the screening criteria in both Refs. 6 and 7, | am certain that these screening
criteria do not address out-of-plane flexure and shear failure modes.

For an aboveground spent fuel pool in which the pool walls (and floor in some cases) are
not supported by soil backfill, it is likely that either out-of-plane flexure or shear will be the
expected seismic failure mode. These walls and floor slab must carry the seismic-induced
hydrodynamic pressure from the water in the pool to their supports by out-of-plane flexure
and shear. It is true that these walls and floor are robust (high strength), but they may not be
as ductile for out-of-plane behavior as they are for in-plane behavior. For an out-of-plane
shear failure to be ductile requires shear reinforcement in regions of high shear.
Furthermore, if large plastic rotations are required to occur, the tensile and compression steel
needs to be tied together by closely spaced stirrups. I question whether such shear
reinforcement and stirrups exist at locations of high shear and flexure in the spent fuel pool
walls and floor. As a result, I suspect that only limited credit for ductility can be taken.

Without taking credit for significant ductility, it is not clear to me that spent fuel pool
walls and floors not supported by soil can be screened at a seismic HCLPF capacity level as
high as 1.2g PSA (equivalent to 0.5g PGA). I am aware of only one seismic fragility analysis
having been performed on such unsupported spent fuel pool walls. That analysis was the
Vermont Yankee spent fuel pool analysis reported in Ref. 5 for which the reported seismic
HCLPF capacity was 0.48g PGA. A single analysis case does not provide an adequate basis
for establishing a screening level for all other cases, particularly when the computed result is
right at the desired screening level. The screening criteria in Refs 6 and 7 are based upon the
review of many cases at more that a dozen plants.

A2B-15


Table 12 Results of Plume Heat Content Sensitivities
(95% evacuation, Surry Population Density)

Case | Release Fraction Plume Mean Consequences (within
Heat 100 miles)
Conten
Cs | Ru Te | Ba Sr Ce La t(MW) | Early Societal Cancer
Fatali- Dose Fatali-
ties (person- ties
rem)
1 1 2x105 | .02 | .002 | .002 | 1x10* | 1x10* | 3.7 1.01 4.54x10° 2,320
45 1 1 02 | .002 | .002 | 1x10 | 1x10* | 3.7 92.2 | 9.50x10® 9,150
47 1 1 02 | .002 | .002 | 1x10® | 1x10* | 83.0 57.3 9.24x10° 9,280
49 1 1 .02 | .002 | .002 | 1x10* | 1x10* | 256.0 18.3 8.24x10° 8,380
46 1 41 .02 | 002 | .002 | 1x10® | 1x10* | 3.7 1.32 6.84x10° 6,430
48° 1 41 02 | .002 | .002 | 1x10* | 1x10* | 83.0 00509 7.28x10° 7,060
50° 1 1 .02 | .002 | .002 | 1x10* | 1x10* | 256.0 .00357 6.96x10® 6,650

*Based on evacuation before release.

Cases 45, 47, and 49 were based on evacuation about an hour after the release began. The staff also
performed calculations based on evacuation beginning three hours before the release begins. Case 46,
which used a ruthenium release fraction of one and evacuation beginning three hours before the release
begins, is shown in the fourth row of Table 12 and was the starting point for these calculations. Then,
Cases 48 and 50 were run with plume heat contents of 83.0 MW and 256 MW, respectively. Increasing
the plume heat content from 3.7 MW to 83.0 MW resulted in a factor-of-300 decrease in the early
fatalities and a small increase in the long-term consequences. Increasing the plume heat content from
83.0 MW to 256 MW resulted in a small decrease in the early fatalities and a small decrease in the long-
term consequences.

Plume Spreading:

MACCS uses a Gaussian plume model with the amount of spreading determined by the parameters a,
and o,, where y is the cross-wind direction and z is the vertical direction. In NUREG/CR-6244,
phenomenological experts provided updated values for o, and o,. However, the experts did not provide
single values of these parameters. Instead, they provided probability distributions. To assess the
sensitivity of spent fuel pool accident consequences to the updated values for o, and o., Sandia National
Laboratories performed MACCS calculations using values for o, and ¢, randomly selected from the
experts distributions.* These MACCS calculations were based on Cases 11 and 14 (see Table 3), which
use the Surry population density and a ruthenium release fraction of one. Case 11 has evacuation
beginning about an hour after the release begins, while Case 14 has evacuation beginning three hours
before the release begins. A total of 300 MACCS runs were performed to generate distributions of early
fatalities, population dose, and cancer fatalities. The results of these MACCS runs are shown in Tables
143 and 14. For the late evacuation case, Case 11, the 50" percentile and mean results using
NUREG/CR-6244 plume spreading are lower for early fatalities and higher for societal dose and cancer
fatalities. The same trend is seen for the early evacuation case, Case 14. Overall, the effect of the
plume spreading model on offsite consequences is not large.

Table 13 Results of Plume-Spreading Model Sensitivity for Case 11
(99.5% evacuation, Surry Population Density)


structural detailing at connections. Similarly, concrete frame reviews should concentrate on the
adequacy of the reinforcement detailing and embedment.

Failure of the structural steel superstructure should be evaluated for its potential impact on
the ability of the spent fuel pool to continue to successfully maintain its water inventory for
cooling and shielding of the spent fuel.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 6:
Requirement: Verify the Adequacy of Spent Fuel Pool Penetrations

Basis: The seismic and structural adequacy of any spent fuel pool (SFP) penetrations whose
failure could result in the draining or syphoning of the SFP must be evaluated for the forces and
displacements resulting from a beyond-design-basis seismic event in the 0.45-0.5g pga range.
Specific examples include SFP gates and gate seals and low elevation SFP penetrations, such
as, the fuel transfer chute/tube and possibly piping associated with the SFP cooling system.
Failures of any penetrations which could lead to draining or syphoning of the SFP should be
considered.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

ftem 7:
Requirement: Evaluate the Potential for Impacts with Adjacent Structures

Basis: Structure-to-structure impact may become important for earthquakes significantly above
the SSE, particularly for soil sites. Structures are usually conservatively designed with rattle
space sufficient to preclude impact at the SSE level but there are no set standards for margins
above the SSE. In most cases, impact is not a serious problem but, given the potential for
impact, the consequences should be addressed. For impacts at earthquake levels below

0.5g pga, the most probable damage includes the potential for electrical equipment malfunction
and for local structural damage. As cited previously, these levels of damage may be found to be
acceptable or to result in the loss of SFP support equipment. The major focus of this impact
review is to assure that the structure-to-structure impact does not result in the inability of the
SFP to maintain its water inventory.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 8:
Requirement: Evaluate the Potential for Dropped Loads

Basis: A beyond-design-basis seismic event in the 0.45-0.5g pga range has the potential to
cause the structural collapse of masonry walls and/or equipment supports systems. If these

A2B-9


Comment #26: What measures have been taken to help minimize fuel handler error in
postulated SFP accident scenarios?

Response: Having procedures in place helps reduce that chance of human errors, especially
under stressful conditions such as during a severe accident. The industry has committed to
providing procedures or administrative controls to reduce the likelihood of rapid drain down
events. IDC #2 is credited for ensuring that procedures and training of personnel are to be in
place to ensure that on-site and off-site resources can be brought to bear during an accident.
IDC #3 is credited to have procedures for establishing communication between onsite and
Offsite organizations during severe weather and seismic events. IDC #4 is credited to ensure
that an off-site resource plan will be developed that will include access to portable pumps and
emergency power. IDC #5 is credited to ensure that fuel handlers will have available to them
spent fuel pool instrumentation that monitors spent fuel pool temperature, water level, and area
radiation levels. In addition, SDA #2 calls for procedures and guidance for plant personnel on
onsite and offsite makeup capability. SDA #3 calls for the direct measurement of water level
and temperature. These staff assumptions and industry commitments are discussed in this
report.

Comment #27: The NRC should review the need to place a containment around spent fuel
pools.

Response: The staff evaluated the risk from spent fuel pool operation and from zirconium fires
at operating plants in Generic Issue 82, “Beyond Design Basis Accidents in Spent Fuel Pools.”
NUREG-1353 determined that the risks of spent fuel pool operation and the cost of alterations
did not justify performing any generic backfits at operating plants, including installation of
containment structures. The staff believes that containment structure is not warranted for
decommissioning spent fuel pools. This issue is discussed in Section 4.1.2 as part of
evaluation of defense in depth.

Comment #28: To the extent possible, experimental validation of risk-informed results should
be addressed.

Response: The predictive models used for estimating the risk from spent fuel pools are based
on a wealth of experimentation. Many experiments have been performed in the areas of
human reliability analysis, seismic fragility of equipment, fires, and thermal hydraulics. The
results of the decommissioning SFP risk assessment come from a systematic analytical
modeling of the SFP and its support systems at a “typical” decommissioning site. The model of
the SFP and its support systems was based on plant-specific visits made by the staff. The staff
used failure rates of support system equipment based on existing large databases of
equipment failure rates. Human error rates were developed by the staff with help from experts
at Idaho National Engineering and Environmental Laboratory. Heavy load drops were based
on modeling performed for NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants,
Resolution of Generic Technical Activity A-36” with additional sources of data from U.S. Navy
crane experiences, Waste Isolation Plant Trudock Crane System experience, and data
supplied by NEI. The effects of aircraft crashes were analyzed using Department of Energy
models and generic aircraft crash data. (See Appendix 2d)

Comment #29: The staff's report is misleading when it states that there is about a factor-of-two

Appendix 6 A6-9


In my judgement, it will be necessary to have either seismic fragility or seismic margin
HCLPF computations performed on at least six different aboveground spent fuel pools with
walls not supported by soil before out-of-plane flexure and shear HCLPF capacity screening
levels can be established for such spent fuel pools.

3.2 Spent Fuel Pool Racks

I don't know whether a gross structural failure of the spent fuel racks is of major concern.
This is a topic outside of my area of expertise. However, if such a failure is of concern, no
seismic HCLPF capacity screening criteria is available for such a failure. The screening
criteria of Refs. 6 and 7 were never intended to be applied to spent fuel pool racks. Since I
have never seen a seismic fragility or seismic margin HCLPF capacity evaluation of a spent
fuel pool rack, I have no basis for deciding whether these racks can be screened at a seismic
HCLPF capacity as high as 1.2g PSA (equivalent to 0.5g PGA).

3.3 Seismic Level 2 Screening Requirements

In order to screen at a seismic HCLPF capacity of 1.2g PSA (0.5g PGA), the Level 2
screening criteria for concrete walls and diaphrams requires that such walls and diaphrams
essentially comply with the ductile detailing and rebar development length requirements of
either ACI 318.71 or ACI 349.76 or later editions. It is not clear to me how many CEUS
spent fuel pool walls and floors essentially comply with such requirements since earlier
editions of these codes had less stringent requirements. Therefore, it is not clear to me how
many spent fuel pool walls and floors can actually be screened at Seismic Level 2 even for
in-plane flexure and shear failure mode.

4. Seismic Risk Associated With Screening Level 2
4.1 Simplified Approaches for Estimating Seismic Ri: iven the HCLPF Capacit

As mentioned in Section 2, the seismic risk of failure of the spent fuel pool can be
estimated by either rigorous convolution of the seismic fragility and the seismic hazard,
or by a simplified approximate method. The simplified approximate method defined by
Eqn. (3) was used in Ref. 1. However, as also mentioned in Section 2, this approximate
method understates the seismic risk by a factor of 2 to 4 for typical CEUS hazard
estimates.

Ref. 10 presents an equally simple approach for estimating the seismic risk of
failure of any component given its HCLPF capacity Cycipr and a hazard estimate. This
approach tends to introduce from 0% to 25% conservative bias to the computed seismic
risk when compared with rigorous convolution. Given the HCLPF capacity Cycipr this
approach consists of the following steps:

Step 1: Estimate the 10% conditional probability of failure capacity Cy, from:
Cro, = FeCucupr (6)
F, = e048

A2B-16


decommissioning plants due to sabotage is low enough to justify modification of safeguards
requirements for SFPs at decommissioning plants.

In the past, decommissioning licensees have requested exemptions from specific regulations in
10 CFR 73.55 on the basis of the reduced number of target sets susceptible to sabotage attacks
and the consequent reduced hazard to public health and safety. Limited exemptions have been
granted on this basis. The risk analysis in this study does not refute the reduced target set
argument; however, the analysis does not support the assertion of a lesser hazard to public
health and safety, given the possible consequences of sabotage-induced uncovery of fuel in the
SFP when a zirconium fire potential exists. Further, the risk analysis in this study did not
evaluate the potential consequences of a sabotage event that could directly cause offsite fission
product dispersion, for example, a vehicle bomb driven into or otherwise significantly damaging
the SFP, even after a zirconium fire was no longer possible. However, this study supports a
regulatory framework that relieves licensees from selected requirements in 10 CFR 73.55 on the
basis of target set reduction when all fuel has been placed in the SFP.

As a result of the conclusions from this study, the bases for previous exemptions for defueled
facilities, the devitalization of the spent fuel pool at operating reactors, and certain concerns at
ISFSIs may need to be reconsidered. This is due to differences in the findings relative to the
specific periods of time historically used for the devitalization of spent fuel pools at operating
reactors and certain operational concerns and potential vulnerabilities at decommissioning sites.

The risk estimates contained in this study are based on accidents initiated by random equipment
failures, human errors, or external events. PRA practitioners have developed and used
dependable methods for estimating the frequency of such random events. By contrast, this
analysis and PRA analyses in general do not include events due to sabotage. No established
method exists for estimating the likelihood of a sabotage event. Nor is there a method for
analyzing the effect of security provisions on that likelihood. Security regulations are designed
and structured to prevent sabotage on the assumption that the design-basis threat could occur
at commercial nuclear power plants without assessing the actual probability or consequences.

The technical information contained in this study shows that the consequences of a zirconium
fire would be high. Moreover, the risk analysis could be used effectively to help determining
priorities for, and details of, the security capability at a plant. However, no information in the
analysis bears on the level of security necessary to limit the risk from sabotage events. Those
decisions will continue to be made by a analytical assessment of the level of threat and the
difficulty of protecting a specific facility.

4.2.3. Insurance

In accordance with 10 CFR Part 140 (Ref. 10), each 10 CFR Part 50 licensee is required to
maintain public liability coverage in the form of primary and secondary financial protection. This
coverage is required to be in place from the time unirradiated fuel is brought onto the facility site
until all of the radioactive material has been removed form the site, unless the Commission
terminates the Part 50 license or otherwise modifies the financial protection requirements under
Part 140. On March 17, 1999, the staff proposed to the Commission that insurance indemnity
requirements for permanently shutdown reactors be developed in an integrated, risk-informed

4-15


Comments #83 and 84: (A) Discuss protection of plant workers, particularly for less severe
accidents such as poo! uncovery without a zirconium fire. (B) The draft report should be
revised to include credible hazards to plant workers at permanently closed plants.

Response: This technical study was limited to accidents involving the draining a
decommissioning plant spent fuel pool. For on-site hazards, the staff believes that existing
regulatory requirements adequately address the need for emergency plans to consider
protective actions and a means for controlling exposures in an emergency for emergency
workers. For example, 10 CFR Part 20 establishes standards for radiation protection for onsite
workers and the public, and 10 CFR 50.47 (b)(10) and (11) establish protective actions and
exposure control regulations for emergency workers. Nuclear power plant licensees are also
subject to regulations for byproduct material under 10 CFR Part 30. Emergency plans under
Section 30.32 require identification of accidents and means for mitigation, including the
protection of onsite workers. Additionally, OSHA and NRC regulations require safety training
and education, including safe handling and use of poisons, caustics, flammable liquids, gases
and toxic materials; radiation protection; and occupational safety.

Although this study does not directly assess accidents or hazards that could occur to plant
personnel, measures for worker safety were included. For example, IDC #8 calls for remote
alignment of the water makeup source to the SFP without requiring entry to the refueling floor,
which prevents workers and other accident responders from entering a potential radiation area.

Comment #85: What will the NRC staff do to protect plant workers and the public from spent
fuel pool risks at permanently closed plants and operating plants before the industry
commitments and staff assumptions are implemented?

Response: The analysis for this study reflects practices already in place. The staff visited four
decommissioning sites as part of the preparation for developing the risk assessment of
decommissioning spent fuel pools. The insights from those visits include that the facilities
appeared to have been staffed by well trained and knowledgeable individuals with significant
nuclear power plant experience. Procedures were in place for dealing with routine losses of
inventory. Fuel handlers appeared to know whom to contact off-site if difficulties arose with the
SFP. The staff recognized that these attributes were not required by any NRC regulations nor
suggested in any NRC guidance for decommissioning sites. The IDCs and SDAs are an
attempt to increase the assurance that plant personnel will continue to be knowledgeable of off-
site resources and have good procedures available to them.

The staff believes that current worker safety regulations adequately protect workers. The
regulations for the protection of workers are the same at decommissioning plants as at
operating plants, such as 10 CFR 20 for standards for protection against radiation. Several
other comments in this appendix also address worker safety regulations.

Comment #86: The NRC should determine the proper methods of extinguishing a possible
zirconium fire.

Response: At the present time, the state-of-art for zirconium fire experiments has not
advanced to determined the various methods for extinguishing. Additional research would
need to be performed to investigate acceptable methods, required quantities of fire-fighting
materials, conditions of use, and guidelines. Due to the low probability of the event, this
research is not recommended at this time.

Appendix 6 A6-28


APPENDIX 4B
Memo to Gary M. Holahan from Farouk Eltawila dated 10/26/00 re:

RADIOLOGICAL CONSEQUENCES OF SPENT FUEL POOL ACCIDENT OCCURRING UP
TO 10 YEARS AFTER FINAL REACTOR SHUTDOWN

Appendix 4B A4B-1


APPENDIX 1.4
THERMAL-HYDRAULICS ANALYSIS OF SPENT FUEL POOL HEATUP

1.0 INTRODUCTION

Spent fuel heatup analyses involving postulated loss of coolant or loss of cooling accidents were
performed to support the decommissioning rulemaking effort. The staff developed spent fuel
heatup models for boiling water reactors (BWRs) and pressurized water reactors (PWRs) that
take into account the decay power and configuration of the fuel in the spent fuel pool (SFP) and
the air flow in the building surrounding the SFP. Discussions in this Appendix include an
explanation of how the thermal-hydraulic estimates of heatup times are used in the rest of the
report. Important assumptions in the analysis (e.g., oxidation rates, ventilation rates, ignition
temperatures, and fuel burnup), are discussed in the context of estimating the heatup and boiloff
times for the SFP inventory, and heatup and ignition timing of fuel cladding to a zirconium fire.
Limitations in the state-of-the-art in performing these calculations are also discussed. Technical
bases are given in Appendix 1.B. for spent fuel cladding temperature criteria used to estimate
when significant fission product releases occur in decommissioning plant SFP accidents.

The time it takes to uncover spent fuel because of pool inventory boiloff is an input to the risk
assessment. This information is used to estimate human error rates and repair time available
for fuel handlers faced with inoperative equipment. Calculations for heatup and boiloff of SFP
inventory involve heatup of the pool water to boiling followed by boil down of the inventory to
within 3 feet of the top of the spent fuel. The time it takes to heatup fuel cladding to zirconium
fire/fission product release temperatures is important in the establishment of consequence
estimates regarding how evacuations would proceed following either loss of inventory or cooling
to an SFP.

1.1 SFP Inventory Heatup and Boiloff

The staff conducted a thermal-hydraulic assessment of the SFP for various scenarios involving
loss of pool cooling and loss of inventory in support of the risk assessment. These calculations
resulted in the estimates of heatup and boiloff times for an SFP, which are displayed in

Table A1-1. These estimates are straight forward energy balance calculations based on
representative decay heat levels, representative volumes in the SFPs, and standard water
properties. The end state used for these accident sequences was an SFP water level 3 feet
above the top of the fuel. This simplified end state was used because recovery below this level,
given failure to recover before reaching this level, was judged to be unlikely given the significant
radiation field in and around the SFP at lowered water levels. The simplified end state provides
a slightly conservative, but adequate measure to determine time frames important to human
error and recovery estimates. It also greatly simplifies the analysis by eliminating the need to
accurately model the complex heat transfer mechanisms and chemical reactions that are
occurring in the fuel assemblies as they are being slowly uncovered.

Appendix 1A A1A-1


Table 3.7-1 - Consequences of an SFP Accident With a High Ruthenium Source Term (per

event)
Mean Consequences for High Ruthenium Source Term (Surry population,
95% evacuation)
Individual Risk*
Societal Dose Individual Risk" | of Latent Cancer
Time After (p-rem within 50 | of Early Fatality Fatality (within
‘Shutdown Early Fatalities miles) (within 1mile) 10 miles)
Late Evacuation

30 days. 192 2.37x10" 4.43x10? 8.24x107
90 days. 162 2.25x10" 4.19x10? 8.20x10?
1 year 7 1.93x10" 3.46x10? 8.49x107

2 years 19 1.69x10" 2.57x10? 8.42x10?
5 years 1 1.45x10" 8.96x107 7.08x107
10 years - 1.34x10" 4.68x107 6.39x107

Early Evacuation

30 days 7 1.35x107 2.01x107 4.79x10*
90 days 4 1.28x10" 1.87x10* 4.77x107
1 year 1 1.12x10? 1.50x10° 4.33x10*
2 years - 9.93x10* 1.12x10% 3.70x10*
5S years - 8,69x10° 3.99x10* 2.93x10*
10 years - 8.13x10® 2.05x10* 2.64x10°

* Conditional on event - Total frequency for all events is shown in Table 3.1 as less than 3x10° per year.

3-29


Figure 4.2 Fire initiating event tree

if FREEVENT | CONTROLROOM ‘OTHER ‘SFPC SYSTEM | OPERATOR RECOVERY | ]
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A similar approach was used to correlate the ignition frequency for "other" to a value appropriate
for the SFP facility. Intake structures typically have several pumps (e.g., circulating water,
service water, screen wash, fire, etc.) as well as peripheral equipment. For this analysis, all
ignition frequency associated with the "other" category was apportioned to pumps. The number
of pumps in the typical intake structure was estimated to be 10 (again, engineering judgement
based on piant walk-downs). Therefore, the fire ignition frequency for "other" equipment at the
SFP facility is estimated to be (5/10)(3.2E-3/yr) = 1.6E-3/yr.

The contribution of ignition sources, identified as "plant-wide" sources in the FIVE document, to
the ignition frequency of the SFP facility is considered to be negligible. Large ignition source
contributors such as elevator motors, dryers, and MG sets do not exist in the spent fuel facility.
Additionally, spontaneous cable fires are expected to be a negligible contributor because of the
minimal amount of energized electrical cable. The facility configuration is expected to be stable,
negating the need for modification and fabrication work requiring welding and cutting.

Appendix 2A A2A-25


(cognition and execution) given the lack of detailed knowledge about expected plant
practices and designs. The PSFs used in the model allowed the impact of the NEI
commitments and additional staff assumptions to be incorporated explicitly into the
evaluation.

3.3. Other Inputs to the Risk Model

A variety of other inputs were required for this PRA, including generic configuration data used in
the fault tree models, radiological calculations, and timing calculations. Initiating event
frequencies and generic reliability data were derived from other studies sponsored by the NRC.
The times available for operator actions are based on calculations of the time it would take for
bulk boiling to begin in the pool, or on the time it takes for the level in the pool to fall to the level
of the fuel pool cooling system suction, or to a height of approximately 3 ft above the fuel, as
appropriate to the definition of the corresponding human failure event.

It takes a relatively long time to uncover the fuel if the initiating event does not involve a
catastrophic failure of the pool. This is because of the large amount of water in an SFP, the
large specific heat of water, and the large latent heat of vaporization for water. Calculations that
were used in the June 1999 and February 2000 study for a typical-sized SFP yield the results in
Table 3.1. Subsequently the staff determined that it had made a mistake in the assumed heat
load on the generic SFP. Current estimates of time to bulk boiling are actually about twice those
given in Table 3.1 below. The staff debated whether to redo the human reliability analysis
estimates assuming the longer periods. It was determined that the credit given for fuel handler
recovery was already so great that it would be difficult to estimate the numerical benefit of the
additional time. Rather, it can be inferred that the uncertainties of whether the absolute value of
the recovery estimates are really so large have been reduced. In addition, the numerical
estimates for the sequences that are affected by these longer recovery times are already so low
that they contribute very little to the overall risk estimates, which are dominated by seismic
events, heavy load drop, and loss of offsite power because of extreme weather that are not as
strongly affected by fuel handler error.

The bulk boiling and boil-off results are based on the following assumptions:

no heat losses

atmospheric pressure

Heat of vaporization h, = 2258 kJ/kg

base pool heat load for a full pool of 2 MW

core thermal power of 3293 MW

typical poo! size (based on Tables 2.1 and 2.2 of NUREG/CR-4982, Ref. 10)
* typical BWR pool is 40' deep by 26’ by 39'

* typical PWR pool is 43' deep by 22’ by 40’

ee eevee

Appendix 2A A2A-9


Figure 5 BWR Low Density Storage Rack Eigenvalue Following Compressive/Expansion

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Appendix 3 A3-18


primary or the secondary system, in which case the response required is simply to start the
redundant pump. However, it could also be a more significant failure, such as a pipe break or a
heat exchanger blockage. To simplify the model, it has been assumed that a repair is
necessary. While this is conservative, it does not unduly bias the conclusions of the overall
study.

If the loss of cooling was detected via the control room alarms, the staff has the full 33 hours in
which to repair the system. Assuming that it takes at least 16 hours before parts and technical
help arrive, then the operators have 17 hours (33 hours less 16 hours) to repair the system.
Failure to repair the SFPC system event is modeled as HEP-COOL-REP-E. This case is
modeled by fault tree LOC-OCS-U.

If the loss of cooling was discovered during walk-downs, it has been conservatively assumed
the operator has only 9 hours available (allowing 24 hours before loss of cooling was noticed).
Since it is assumed that it takes at least 16 hours before technical help and parts arrive, it is not
possible that the SFPC system can be repaired before the bulk boiling would begin. Failure to
repair the SFPC system event is modeled as HEP-COOL-REP-L. This case is modeled by fault
tree LOC-OCS-L.

4.1.4.2 Relevant Assumptions
. The operators will avoid using raw water (e.g., water not chemically controlled) if possible.
Therefore, the operators are assumed to focus solely on restoration of the SFP cooling

system in the initial stages of the event.

. If the loss of cooling was detected through shift walk-downs, then 24 hours are
(conservatively) assumed to have passed before discovery.

. It takes 16 hours to contact maintenance personnel, diagnose the cause of failure, and get
new parts.

. Mean time to repair the SFP cooling system is 10 hours.

. Operating staff has received formal training and there are administrative procedures to
guide them in initiating repair (NEI commitment no. 8).

. Repair crew is different than the onsite operators.

Appendix 2A A2A-18


secondary structural failures could result in the accidental dropping of heavy loads which are
always present (i.e. not loads associated with cask movements) into the SFP, then the
consequences of these drops must be considered. As in previous evaluations, the focus of the
drop consequence analyses should consider the possibility of draining the SFP. Additionally,
the evaluation should evaluate the consequences of any resulting damage to the spent fuel or to
the spent fuel storage racks.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 9:

Requirement: Evaluation of Other Failure Modes

Basis: Experienced seismic engineers should review the geotechnical and structural design
details for the specific site and assure that there are not any design vulnerabilities which will not
be adequately addressed by the review areas listed above. Soil-related failure modes including
liquefaction and slope instability should be screened by the approaches outlined in Reference 1
(Section 7 & Appendix C).

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 10: Potential Mitigation Measures

Although beyond the scope of this seismic screening checklist, the following potential
mitigation measures may be considered in the event that the requirements of the seismic
screening checklist are not met at a particular plant.

a.) Delay requesting the licensing waivers (E-Plan, insurance, etc.) until the plant specific
danger of a zirconium fire is no longer a credible concern.

b.) Design and install structural plant modifications to correct/address the identified areas of
non-compliance with the checklist. (It must be acknowledged that this option may not be
practical for significant seismic failure concerns.)

c.) Perform plant-specific seismic hazard analyses to demonstrate that the seismic risk
associated with a catastrophic failure of the pool is at an acceptable level. (The exact
“acceptable” risk level has not been precisely quantified but is believed to be in the range of
1x10° per year.)

We believe that use of the checklist and determination that the spent fuel poo! HCLPF is

sufficiently high will assure that the frequency of fuel uncovery from seismic events is less than
or equal to 1x10°* per year.

A2B-10


regular, (once per shift) basis is important to compensate for potential failures to the
instrumentation monitoring the status of the pool. The analysis has also assumed that the
procedures and/or training are explicit in giving guidance on the capability of the fuel pool
makeup system, and when it becomes essential to supplement with alternate higher volume
sources. The analysis also assumed that the procedures and training are sufficiently clear in
giving guidance on early preparation for using the alternate makeup sources.

Table 4.1 Basic Event Summary for the Loss of Cooling Event Tree

. ones Basic Event
Basic Event Name Description Probability
lE-LOC Loss of SFP cooling initiating event 3.0E-3

_ Operators fail to respond to a signal
HEP-DIAG-ALARM indication in the control room 3.0E-4
Operators fail to observe the loss of
HEP-WLKDWN-LSFPC. cooling in walk-downs (independent 1.0E-5
case)
Operators fail to observe the loss of
HEP-WLKDWN-DEPEN cooling in walk-downs (dependent case) 5.0E-2
HEP-COOL-REP-E Repair crew fails to repair SFPC system 1.8E-1
HEP-COOL-REP-L Repair crew fails to repair SFPC system 1.0
Operators fail to diagnose need to
HEP-RECG-FWSTART start the firewater system 2.0E-5
Operators fail to diagnose need to start 5.0E-2
HEP-RECG-FWSTART-L | firewater system - dependent case °

EW. Operators fail to start firewater pump
HEP-FW-START and provide alignment 1.0E-5

EW. ¥ Repair crew fails to repair firewater _
HEP-FW-REP-DEPEN system - dependent case 5.0E-2

Operators fail to provide alternate
HEP-INV-OFFSITE sources of cooling from offsite 5.0E-2
FP-2PUMPS-FTF Failure of firewater pump system 6.7E-4
SPC-LVL-LOP Local faults leading to alarm channel 2.0E-3
failure

4.2 Internal Fire Event Tree

This event tree models the loss of SFP cooling caused by internal fires. Given a fire alarm, the
operator will attempt to suppress the fire, and then attempt to re-start SFP cooling given that the
SFP cooling system and offsite power feeder system have not been damaged by the fire. In the
unlikely event that the operator fails to respond to the alarms or is unsuccessful in suppressing
the fire, it is assumed that the SFPC system will be damaged to the extent where repair will not
be possible. The operator then has to provide alternate cooling and inventory makeup - either
using the site firewater system or by calling upon offsite resources. Figure 4.2 shows the
Internal Fire event tree sequence progression.

Appendix 2A A2A-23


Table A2c-3 WIPP Evaluation for Failure to Secure Load (improper rigging estimate)

Symbol | HEP Explanation of error Source of HEP
(NUREG/CR-1278)

A, 3.75x10° | Improperly make a connection, including failure to | Table 20-12 Item 13
test locking feature for engagement Mean value (0.003, EFTM = 3)

B, 0.75 The operating repeating the actions is modeled to | Table 20-21 Item 4(a)
have a high dependency for making the same High dependence for different
error again. It is not completely independent pins. Two opportunities (the
because the operator moves to the second lifting second and third pins) to repeat
leg and must physically push the locking balls to the error is modeled as
insert the pins 0.5+(1-0.5)*0.5 = 0.75

Cc, 4.25x10° | Checker fails to verify proper insertion of the Table 20-22 Item 9
connector pins, and that the status affects safety Mean vatue (0.001, EF = 3)
when performing tasks

D, 0.15 Checker fails to verify proper insertion of the Table 20-21 Item 3(a)
connector pins at a later step, given the initial Moderate dependency for
failure to recognize error. Sufficient separation in second check
time and additional cues to warrant moderate
rather than total or high dependency.

F, 5.2x107 Failure rate if first pin improperly connected C,*D,

a 0.99625 Given first pin was improperly connected

Ad 3.75x10°% | improperly make a connection, including failure to | Table 20-12 Item 13
test locking feature for engagement Mean value (0.003, EF = 3)

B, 0.5 The operating repeating the actions is modeled to | Table 20-21 Item 4(a)
have a high dependency for making the same High dependence for different
error again. It is not completely independent pins. Only one opportunity for
because the operator moves to the second lifting error (third pin)
leg and must physically push the locking balls to
insert the pins

Cc, 4.25x10° | Checker fails to verify proper insertion of the Table 20-22 Item 9
connector pins, and that the status affects safety Mean value (0.001, EF = 3)
when performing tasks

D, 0.15 Checker fails to verify proper insertion of the Table 20-21 Item 3(a)
connector pins at a later step, given the initial Moderate dependency for
failure to recognize error. Sufficient separation in second check
time and additional cues to warrant moderate
rather than total or high dependency.

F, 3.5x107 Failure rate if first pin improperly connected a, *A,* Bo * C2 *D2

F, 8.7x107 Total failure because of human error F1+F2

(1) Note: The EF (error factor) is the 95" percentile/50" percentile (median). For an EF of 3,

the mean-to-median multiplier is 0.8.

A2C-8



Table 3
Seismic Risk Associated With Screening Level 2

Cucipr= 1.2g Peak Spectral Acceleration

Annual Seismic-Induced
Site Probability of Failure P,
Number (to be muitiplied by 10°)
LLNL93 Hazard | EPRI89 Hazard
36 13.6 0.14
18 8.3 1.9
25 6.6 0.57
8 5.5 0.21
43 45 0.12
59 44 *
21 42 *
62 41 *
27 2.9 0.38
49 2.8 0.27
40 2.5 0.10
16 2.5 0.14
38 2.3 0.21
63 2.2 0.06
54 2.2 0.26
19 1.8 0.17
32 1.8 0.17
28 17 0.04
4 1.6 *
50 1.5 0.20
44 1.5 *
20 1.5 0.55
31 1.4 0.06
39 1.4 0.14
14 1.3 0.60
13 1.3 0.33
Not Available

A2B-22



4.4.4 Top Event OMK - Operator Recovery Using Makeup Systems

4.4.4.1 Event Description and Timing

This top event represents the failure probability of the firewater pumps. If offsite power is
recovered then the fault tree LP2-OMK-U represents this top event. In this case, the operators
have both electric and diesel firewater pumps available. If offsite power is not recovered then
fault tree LP2-OMK-L represents this top event. In this case, the operator has only the diesel
firewater pump available.

4.4.4.2 Relevant Assumptions

. It is assumed that the procedures guide the operators to wait until it is clear that SFP
cooling cannot be reestablished (e.g., using cues such as the level drops to below the
suction of the cooling system or the pool begins boiling) before using alternate makeup
sources. Therefore, they have 88 hours to start a firewater pump.

. Because of the severe weather, if one or both pumps fail to start or run, it is assumed that
it takes another four to five shifts (48 hours) to contact maintenance personnel, perform
the diagnosis, and get new parts. Therefore, the operator would have 40 hours (88 hours
less 48 hours) to perform repairs.

e There is a means to remotely align a makeup source to the SFP without entry to the refuel
floor, so that make-up can be provided even when the environment is uninhabitable
because of steam and/or high radiation (NEI commitment no.8).

. Repair crew is different than onsite operators.

. Repair crew will focus his recovery efforts on only one pump

. On average, it takes 10 hours to repair a pump ff it fails to start and run.

. Both firewater pumps are located in a separate structure or protected from the potential
harsh environment in case of pool bulk boiling.

. Maintenance is performed per schedule on diesel and electric firewater pumps to maintain
operable status.

« Operators haves received formal training on relevant procedures.

4.4.4.3 Quantification

Human Error Probabilities

The fault tree LP2-OMK-U has five operator actions, and LP2-OMK-/ has three. Two of the
events are common. HEP-RECG-FWST-SW represents the failure of the operator to recognize

the need to initiate firewater as an inventory makeup system. This event was quantified using

Appendix 2A A2A-47


Table 4.4 Basic Event Summary for Severe Weather Loss of Offsite Power Event Tree

run

Basic Event Name Description Basic Event Probability

1E-LP2 LOSP event because of 1.1E-02
severe-weather-related causes

HEP-DIAG-SFPLP2 Operators fail to diagnose loss of SFP 1.0E-5
cooling because of loss of offsite
power

HEP-RECG-DEPEN Failure to recognize need to cool pool 5.0E-2
given prior failure

HEP-SFP-STR-LP2 Operators fail to restart and align the 5.0E-4
SFP cooling system once power is
recovered

HEP-RECG-FWST-SW Operators fail to diagnose need to 1.0E-4
start the firewater system

HEP-FW-START-SW Operators fail to start firewater pump 1.0E-3
and provide alignment

HEP-FW-REP-DEPSW Repair crew fails to repair firewater 7.0E-2
system

HEP-FW-REP-NODSW Repair crew fails to repair firewater 1.8E-2
system

HEP-INV-OFFST-SW Operators fail to provide alternate 8.0E-2
sources of cooling from offsite

REC-OSP-SW Recovery of offsite power within 2.0E-2
24 hours

SPC-CKV-CCF-H Heat exchanger discharge check 1.9E-5
valves - CCF

SPC-CKV-CCF-M SFP cooling pump discharge check 3.2E-5
valves - CCF

SPC-HTX-CCF SFP heat exchangers ~ CCF 1.9E-5

SPC-HTX-FTR SFP heat exchanger cooling system 24E-4
fails

SPC-HTX-PLG Heat exchanger plugs 2.2E-5

SPC-PMP-CCF SFP cooling pumps - common cause 5.9E-4
failure

SPC-PMP-FTF-1 SFP cooling pump 1 fails to start and 3.9E-3

Appendix 2A

A2A-51



reduction in prompt fatalities if the accident occurs after one year instead of thirty days. The
real insight should be that compared to operating plants, the absolute value of prompt fatalities
from zirconium fires at SFPs is a couple of orders of magnitude lower. In fact, the report does
not justify a one-year delay in eliminating off-site emergency preparedness. Prompt fatalities
are sufficiently reduced one month after reactor shutdown to support eliminating off-site
emergency preparedness.

Response: This report provides the staff's re-evaluation of the risk of spent fuel pool accidents,
as well as the effectiveness of emergency preparedness at decommissioning plants. The staff
has evaluated the consequences at several times after shutdown at times less than one year.

Comment #30: The human error probabilities (HEPs) used for the operator action “Operator
Recovery Using Off-Site Sources” are too conservative.

Response: The HEPs for recovery using off-site sources were quantified with the assumption
that the fuel handlers/plant operators will initially attempt to mitigate the upset condition using
in-house resources, and having failed this, attempt recovery using off-site sources. This was
based on input obtained from licensees during public meetings on this subject, and on the
assumption that fuel handlers will initially avoid using raw water (i.e., water not chemically
controlled) when possible. It was however assumed that licensee procedures and training are
in place to ensure that off-site resources can be brought to bear (IDCs # 2 and 4), and that
these procedures explicitly state that if the water level drops below a certain level (e.g., 15 feet
below normal level), licensee personnel must initiate recovery using off-site sources (SDA # 4)
The probability of this event was quantified under the assumption that there is a low
dependence with preceding fuel handler failures. Given that the event is always coupled with
other fuel handler failures, it would, in the staff's opinion, be inappropriate to argue for zero
dependence. When looked at in the context of the complete cutsets, it can be seen that the
likelihood of failure to respond to any of the initiating events (excluding seismic and heavy load
drops) where meaningful responses are possible is indeed low, as is evident from the low
sequence frequencies.

Comment #31: Is it realistic to assume “good communication” with off-site emergency
organizations once the plant is shutdown and “forgotten”?

Response: As the time after shutdown increases, the decay heat loads decrease and more
time is needed to heat up the pool water and boil off if heat removal were lost. After one year,
the decay heat levels are such that there is at least a week of delay between loss of cooling
and spent fuel uncovery. Even following a seismic or severe weather event, the staff expects
that a utility will be aware of the resources that are available in the area to provide pool cooling
or inventory make up and that the utility will have assured the availability of the resources. In
addition, the utility should have a plan for communicating with suppliers and government
Officials during such emergencies by means that would not be disrupted by such events

(e.g., by portable radio). IDCs #2 and #3 provide assurance that good communication will be
maintained.

Comment #32: Will commitments lead to practices better than current? If not, use historic
data.

Response: It is the staffs expectation that the commitments will in general provide guidance

Appendix 6 A6-10


NUREG-1738

—=

Technical Study of Spent
Fuel Pool Accident Risk
at Decommissioning
Nuclear Power Plants

Manuscript Completed: January 2001
Date Published: February 2001

Prepared by
T.E. Collins, G. Hubbard

Division of Systems Safety and Analysis
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Figure A2e-5 Tornado Exceedance Probability For Each F-scale

1E-04

Tornado Exceedance Probability

Based on NUREG/CR-2944

1E-05

1E-06

1E-07

Exceedance probability (per year)

1E-08

1E-09

FO

F4

F-Scale

F4

FS

—s
Region A

=
Region B
a
Region C
Ea
Region D
HO
USA

Appendix 2E

A2E-11



Table A4-4 Timing of Events

Event Time (sec) Time (hour)
Notification given to offsite emergency response i} 0
Officials

start time of offsite release 2400 7
end time of offsite release 4200 1.2
evacuation begins 7200 2.0

Table A4-5 Evacuation Modeling

Parameter Value

size of evacuation zone 10 miles

sheltering in evacuation zone no sheltering

evacuation direction radially outward

evacuation speed 4 miles/hr

other after evacuee reaches 20 miles from fuel
pool, no further exposure is calculated

After the first week, the pre-accident population in each sector (including the evacuation zone)
is assumed to be present unless the dose to an individual in a sector will be greater than 4 rem
over a period of 5 years. If the dose to an individual in a sector is greater than 4 rem over a
period of 5 years, then the population in that sector is relocated. Dose and cost criteria are
used to determine when the relocated population returns to a sector. The dose criterion is that
the relocated population is returned at a time when it is estimated that an individual's dose will
not exceed 4 rem over the next 5 years. The actual population dose is calculated for exposure
for the next 300 years following the population’s return.

Offsite Consequence Results

MACCS calculations for a decommissioned reactor for accidents occurring 30 days, 90 days,
and 1 year after final shutdown were performed to assess the magnitude of the decrease in the
offsite consequences resulting from extended decay before the release. These calculations
were performed for a Base Case along with a number of sensitivity cases to evaluate the
impact of alternative modeling. These cases are summarized in Table A4-6. The results of
these calculations are discussed below.

Appendix 4 A4-7


4.3.5.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-INV-OFFSITE 5.0E-2

4.3.6 Summary

Table 4.3 presents a summary of basic event probabilities used in the quantification of the
Plant-centered and Grid-related Loss of Offsite Power event tree.

As in the case of the loss of cooling, and fire initiating events, based on the assumptions made,
the frequency of fuel uncovery can be seen to be very low. Again, a careful and thorough
adherence to NEI commitments 2, 5, 8 and 10, the assumption that walk-downs are performed
on a regular, (once per shift) basis is important to compensate for potential failures to the
instrumentation monitoring the status of the pool, the assumption that the procedures and/or
training are explicit in giving guidance on the capability of the fuel pool makeup system, and
when it becomes essential to supplement with alternate higher volume sources, the assumption
that the procedures and training are sufficiently clear in giving guidance on early preparation for
using the alternate makeup sources, are crucial to establishing the low frequency.

Appendix 2A A2A-41


Figure 2.1 Simplified Diagram of Spent Fuel Pool Cooling and Inventory Makeup Systems

A
Makeup { Fi Fuel Poo
Tank | uel Foo!
Makeup Pump
Lé¢ cr
c a
| x |
Hose | i cer | T |
| a xi Filter -——<
lo- rap i : !
| WAY |
eo ms] Lope ox} ;
i rh ~MP-2
| bo ro KA i NA _
! Spent Fuel Pool || nC eas a
| FP) MPA
ni | Fuel Pool
pot —— Cooling Pumps
\
7
_
Reservar rc
WH : i i |
Fire Pi — Ultimate: : ;
venues Heat Sink |
(Air Coolers) | |
Gos er2

Appendix 2A A2A-2


spent fuel pool risk report be extended by 3 months.

Response: The original 45 day comment period ended on April 7, 2000. In a public meeting on
May 9, 2000, NRC managers told the stakeholder that the comment period would be extended
until June 9, 2000.

Comment #107: The NRC should identify and address possible conflicts of interests, and
differing professional opinions as to the use of PRA (probabilistic risk assessment). For
instance, Dr. Hanauer was quoted in a memo to say, “you can make probabilistic numbers
prove anything, by which | mean that probabilistic numbers mean prove nothing.”

Response: It is the policy of the Commission to maintain a working environment that
encourages the employees to make known their best professional judgements even though
they may differ from a prevailing staff view. An objective of this policy is to ensure full
consideration and prompt disposition of differing opinions and views by affording an
independent, impartial review by qualified personnel. The content of the quote is responded to
in this appendix as a separate comment.

Comment #108: The NRC should make references used in the spent fuel pool risk study
available at no cost.

Response: The NRC policy is that all pertinent regulatory information is made available to the
public via the Public Document Room and/or through the Agency Document and Management
System (ADAMS) where this information is available for inspection at no charge. However,
during the period of this study, the NRC took additional actions to provide the stakeholder with
free copies of all routine correspondence and of numerous studies and reports that he
specifically requested. Additionally, the NRC provided free copies of the draft June spent fuel
pool risk study to all interested persons who attended the July 1999 public workshop and to ail
other members of the public who requested it.

Comment #109: Changes to decommissioning regulations should be made on an interim
basis, to be reviewed again at some future date.

Response: The NRC does not plan to issue interim regulations for decommissioning.
Rulemaking is a methodical and deliberately lengthy procedure to ensure that a rule is not
issued without due process. Provisions for public comment as well as independent review
committees afford ample opportunity to examine a rulemaking prior to issuing a new rule. Any
person who believes an NRC regulation is no longer applicable may petition the Commission to
issue rescind, or amend that regulation in accordance with 10 CFR 2.802.

Comment #110: The Draft Study completely sidesteps the question of where all the people
who are relocated will be able to go for the decades that must pass while the land where they
live recovers from radioactive contamination. This issue is graphically illustrated by the
consequences of the Chernobyl accident, which rendered huge land areas uninhabitable and
unsuitable for agriculture for an extended period of time. Finally, the Draft Study fails entirely to
address the social and economic implications of losing the use of thousands of square
kilometers of land for several generations.

Appendix 6 A6-35,


Attachment 1

Appendix 4A Effect of Source Term and Plume-Related Parameters on Consequences
Introduction

Appendix 4 documents the staffs evaluation of the offsite consequences of a spent fuel pool accident
involving a sustained loss of coolant, leading to a significant fuel heatup and resultant release of fission
products to the environment. The objectives of the consequence evaluation were (1) to assess the effect
of one year of decay and (2) to assess the effect of early versus late evacuation because spent fuel pool
accidents are slowly evolving accidents. The staffs evaluation was an extension of an earlier study
performed by Brookhaven National Laboratory (BNL) for spent fuel pools at operating reactors, which
assessed consequences using inventories for 30 days after shutdown."

To perform the evaluation documented in Appendix 4, the staff used the MACCS code (MELCOR
Accident Consequence Code System)’ with fission product inventories for 30 days and 1 year after final
shutdown. The evaluation showed that short-term consequences (early fatalities) decreased by a factor
of two when the fission product inventory was changed from that for 30 days after final shutdown fo that
for one year after final shutdown. It also showed that, at one year after final shutdown, early evacuation
decreased early fatalities by up to a factor of 100. Long-term consequences (cancer fatalities and
societal dose) were unaffected by the additional decay and early evacuation. Representative results for
the Surry population density are shown in Table 1.

Table 1 Representative Results
(99.5% evacuation, Surry Population Density)

Decay Time Prior to Mean Consequences (within 100 miles)
Accident
Early Fatalities Societal Dose Cancer Fatalities
(person-rem)
30 days 1.75 4.77x10° 2,460
4 year 1.01 4.54x10° 2,320
1 year? .0048 4.18x10® 1,990

* Based on evacuation before release.

As noted above, the staffs consequence evaluation was an extension of an earlier consequence
evaluation to gain insight into the effect of one year of decay and of early evacuation. Subsequent
reviews of the staffs consequence evaluation identified issues with the earlier evaluation performed by
BNL in the areas of fractional release from the fuel of each fission product (i.e., fission product source
term) and plume-related parameters. To address these issues, the staff performed additional MACCS
sensitivity calculations which are documented below.

Fission Product Source Term

The Appendix 4 consequence assessment was based on the release fractions shown in Table 2, which
are from the BNL study.’ it also was based on releasing fission products from a number of fuel
assemblies equivalent to 3.5 reactor cores. These release fractions include relatively small release
fractions for the low-volatile and non-volatile fission products.

Table 2 Fission Product Release Fractions from the BNL Study

xenon, iodine | cesium | telluriu strontium

krypton m

lantha-
num

barium | ruthenium cerium



Table A4-16 Comparison with NUREG/CR-6451 Results (short-term consequences)

Dist. Prompt Fatalities
(miles)
NUREG/CR- Case 4 | Case 4
6451 Rerun
0-50 74 18.3 168
0-500 101 18.3 168
Effect of Cesium

Cesium is volatile under severe accident conditions and was previously estimated to be
completely released from fuel under these conditions. Also, the half-lives of the cesium
isotopes are 2 years for cesium-134, 13 days for cesium-136, and 30 years for cesium-137.
Therefore, we performed additional sensitivity calculations on the Base Case to evaluate the
importance of cesium to better understand why the consequence reduction from a year of
decay was not greater. The results of our calculations are shown in Table A4-17. As shown in
this table, we found that the cesium isotopes with their relatively long half-lives were
responsible for limiting the reduction in offsite consequences.

Table A4-17 Mean Consequences for the Base Case with and Without Cesium

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities
Spent Fuel Pool Fatalities (person-Sv)

1 year 0-100 1.01 45,400 2,320
1 year 0-100 0.00 1,460 42
(without cesium)

Conclusion

The primary objective of this evaluation was to assess the effect of extended storage in an
SFP, and the resulting radioactive decay, on offsite consequences of a severe SFP accident at
a decommissioned reactor. This evaluation was performed in support of the generic evaluation
of SFP risk that is being performed to support related risk-informed requirements for
decommissioned reactors. This evaluation showed about a factor-of-two reduction in prompt
fatalities if the accident occurs after 1 year instead of after 30 days. Sensitivity studies showed
that cesium with its long half-life (30 years) is responsible for limiting the consequence
reduction. For the population within 100 miles of the site, 97 percent of the societal dose was
from cesium. Also, this evaluation showed that beginning evacuation three hours before the
release begins reduces prompt fatalities by more than an order of magnitude.

Appendix 4 A4-16


+ The risk to the population in the area near a nuclear power plant of cancer fatalities that
might result from nuclear power plant operation should not exceed one-tenth of 1 percent
(0.1 percent) of the sum of cancer fatality risks resulting from all other causes.

These QHOs have been translated into two numerical objectives as follows:

+ The individual risk of a prompt fatality from all “other accidents to which members of the
U.S. population are generally exposed,” such as fatal automobile accidents, is about
5x10% per year. One-tenth of 1 percent of this figure implies that the individual risk of
prompt fatality from a reactor accident should be less than 5x10” per reactor year.

+ “The sum of cancer fatality risks resulting from all other causes” for an individual is taken to
be the cancer fatality rate in the U.S. which is about 1 in 500 or 2x10* per year. One-tenth
of 1 percent of this risk means that the risk of cancer to the population in the area near a
nuclear power plant due to its operation should be limited to 2x10 per reactor year.

Although the policy statement and related numerical objectives were developed to address the
tisk associated with power operation, the QHOs provide a convenient benchmark for SFP
evaluations. Accordingly, the staff has compared the estimated risks associated with SFP
accidents to the QHOs.

The risks associated with SFP accidents compare favorably with the QHOs. The comparisons,
presented in Section 3.7.3, show that a typical site that conforms with the IDCs and SDAs would
meet the QHOs by one to two orders of magnitude a few months following shutdown and by
greater margins later. The risk comparisons provided in Appendix 4C show that SFP facilities
maintained at or below the recommended pool performance guideline (PPG) of 1X10° per year,
would continue to meet the QHO even with a severe SFP source term. With the exception of
H.B. Robinson (using the LLNL seismic hazard estimates and generic fragilities), all Central and
Eastern U.S. plants which satisfy the IDCs and SDAs (and pass the seismic checklist) will meet
the PPG. Western plants (including San Onofre, Diablo Canyon, and WNP-2) were not included
in the LLNL or EPRI seismic hazard studies and need to demonstrate compliance with the PPG
on a plant-specific basis.

Risk Increases Relative to Regulatory Guide 1.174

The guidelines in RG 1.174 pertain to the core damage frequency (CDF) and large early release
frequency (LERF). For both CDF and LERF, RG 1.174 contains guidance on acceptable values
for the changes that can be allowed due to regulatory decisions as a function of the baseline
frequencies. For example, if the baseline CDF for a plant is below 1x10“ per year, plant
changes can be approved that increase CDF by up to 1x10° per year. If the baseline LERF is
less than 1x10° per year, plant changes can be approved which increase LERF by up to

1x10° per year.

For decommissioning plants, the risk is primarily due to the possibility of a zirconium fire
involving the spent fuel cladding. The consequences of such an event do not equate directly to
either a core damage accident or a large early release as modeled for an operating reactor.
Zirconium fires in SFPs have the potential for significant long-term consequences because

4-3


The consequences for seismic sequences are therefore based on results for the late evacuation
cases in Tables 3.7-1 and 3.7-2. The same reasoning is applied for cases with and without EP
relaxations and for all times after shutdown. The net effect is that EP, as well as relaxations in
EP, do not impact the risk associated with seismic events that result in SFP failure. A sensitivity
study was also done to explore the impact on risk if the seismic event only partially degrades the
emergency response (see Section 4.2.1).

Rapid Draindown Due to Non-Seismic Events
Non-seismically-initiated events leading to rapid draindown are listed in Table 3.7-5. These

events are dominated by cask drop accidents, with the next highest contributor nearly two
orders of magnitude lower.

Table 3.7-5 Frequency of Rapid Draindown Spent Fuel Uncovery Due to Nonseismic
Events
Initiating Event Frequency (per year)
Cask drop 2.0x107
Aircraft impact 2.9x10°
Tornado missile <1.0x10°
Total 2.0x107

Cask drop accidents that lead to catastrophic failure of the SFP include accidents in which the
load is dropped either on the pool floor or on or near the pool wall. Load drops on the pool floor
are more likely to result in complete draindown of the pool and create an air flow path through
the fuel assemblies. Load drops on the pool wall would likely result in residual water in the pool,
which would obstruct air flow. For purposes of consequence assessment, all cask drop
accidents leading to fuel uncovery were assumed to result in a rapid draindown followed by an
SFP fire.

Depending upon the pool failure mode and location, the fuel could be air cooled, or heatup could
be close to adiabatic as a result of air flow blockage. As discussed in Appendix 1A for either
adiabatic or air flow conditions (at 60 GWD/MTU burnup), the time of fission product release
would be about 4 hours for a PWR and 8 hours for a BWR for accidents initiated 1 year following
shutdown. For cases with air cooling, close to 1 day is available after 3 years decay. Even with
adiabatic heatup, 1 day is available after 5 years of decay. At 60 days after shutdown, fission
product release could begin as early as 2 hours after fuel uncovery. The actual time would
depend on reactor type, fuel burnup, fuel rack structure, and other plant-specific parameters, as
discussed in Appendix 1A. The fuel handlers would be immediately aware of a cask drop
accident. It is expected that with procedures that specify the SFP water level at which an
emergency is to be declared, the proper offsite authorities would be promptly informed.

3-38


because fuel fines could be released offsite from fuel with high burnup. While the staff believes
that it is unlikely that fuel fines would be released offsite in any substantial amount, a sensitivity
was performed using a release fraction of 6x10° for lanthanum and cerium to determine
whether such an increase could even impact offsite consequences.

Table A4-3 Release Fractions for a Severe Spent Fuel Pool Accident

Radionuclide Group Release Fractions
NUREG/CR- NUREG/CR-
4982 6451
noble gases 1 1
iodine 1 1
cesium 1 4
tellurium 2x10? 2x10?
strontium 2x10% 2x10°
ruthenium 2x10% 2x10%
lanthanum 1x10* 6x10*
cerium 1x10* 6x10%
barium 2x10 2x10

Modeling of Emergency Response Actions and Other Areas

Modeling of emergency response actions was essentially the same as that used for Surry in
NUREG-1150. The timing of events is given in Table A4-4. Evacuation begins exactly two
hours after emergency response officials receive notification to take protective measures. This
results in the evacuation beginning approximately .8 hours after the offsite release ends. Only
people within 10 miles of the SFP evacuate, and, of those people, .5% do not evacuate.
Details of the evacuation modeling are given in Table A4-5.

People outside of 10 miles are relocated to uncontaminated areas after a specified period of
time depending on the dose they are projected to receive in the first week. There are two
relocation criteria. The first criterion is that, if the dose to an individual is projected to be
greater that 50 rem in one week, then the individual is relocated outside of the affected area
after 12 hours. The second criterion is that, if the dose to an individual is projected to be
greater that 25 rem in one week, then the individual is relocated outside of the affected area
after 24 hours.

Appendix 4 A4-6


10.

41,

12.

13.

14.

15.

16.

17.

18.

19.

20.

Cooper, S. E., et al, "A Technique for Human Error Analysis (ATHEANA),
NUREG/CR-6350, May 1996, USNRC.

Swain, A. D., and Guttman, H. E., "Handbook of Human Reliability Analysis with Emphasis
on Nuclear Power Plant Applications", (THERP), NUREG/CR-1278, August 1983, USNRC.

Byers, J.C., et al., "Revision of the 1994 ASP HRA Methodology (Draft)",
INEEL/EXT-99-00041, January 1999.

Sailor, et. al., "Severe Accidents in Spent Fuel Pools in Support of Generic Issue 82",
NUREG/CR-4982 (BNL-NUREG-52093), July 1987.

U.S. Nuclear Regulatory Commission, "Control of Heavy Loads at Nuclear Power Plants,
Resolution of Generic Technical Activity A-36," NUREG-0612, July 1980.

U.S. Nuclear Regulatory Commission, "Operating Experience Feedback Report -
Assessment of Spent Fuel Cooling,” NUREG-1275, Volume 12, February 1997.

Idaho National Engineering and Environmental Laboratory, "Loss of Spent Fuel Poo!
Cooling PRA: Model and Results,” INEL-96/0334, September 1996.

Electric Power Research Institute, "Fire-Induced Vulnerability Evaluation (FIVE)," EPRI
TR-100370s, April 1992.

OREDA-92 Offshore Reliability Data Handbook, 2nd Edition, 1992.

Atwood, et. al., "Evaluation. of Loss of Offsite Power Events at Nuclear Power Plants: 1980
- 1996," NUREG/CR-5496, November 1998.

U.S. Nuclear Regulatory Commission, "Evaluation of Station Blackout Accidents at
Nuclear Power Plants," NUREG-1032, June 1988.

U.S. Nuclear Regulatory Commission, "Single-Failure-Proof Cranes for Nuclear Power
Plants," USNRC Report NUREG-0554, May 1979.

Appendix 2A A2A-69


release reducing the concentration of activity in the plume of fission products for offsite
consequence analysis.

5.

1.

REFERENCES

Benjamin, A., et al., “Spent Fuel Heatup Following Loss of Water During Storage,”
NUREG/CR-0649, SAND77-1371, Sandia National Laboratories, 1979.

Haste, T., et al., “In-Vessel Core Degradation In LWR Severe Accidents,” EUR 16695 EN,
1996.

Shepherd, I., et al., “Oxidation Phenomena in Severe Accidents- Final Report,”
INV-OPSA(99)-P0008, EUR 19528 EN, 2000.

Evans, E., et al., “Critical Role of Nitrogen During High Temperature Scaling of Zirconium,”
Proc. Symp. Metallurgical Society of AIME, May 1972.

Fischer, S. and M. Gruebelich, “Theoretical Energy Release of Thermites, Intermetallics,
and Combustible Metals,” 24" International Pyrotechnics Seminar, July 1998.

Gauntt, R., et al., “MELCOR Computer Code Manuals Version 1.8.4," NUREG/CR- 6119,
SAND97-2398, July 1997.

Powers, D. et al., “A Review of the Technical Issues of Air Ingression During Severe
Accidents,” NUREG/CR-6218, SAND-0731, September 1994.

Letter from H. M. Chung to J. Flack, dated August 24, 2000.

Leistikow, S. and H. V. Berg, “Investigation under Nuclear Safety Aspects of Zircaloy-4
Oxidation Kinetics at High Temperatures in Air,” Second Workshop of German and Polish
Research on High Temperature Corrosion of Metals, Juelich, Germany, December 2-4,
1987.

Appendix 1B A1B-8


section processing, and depletion calculations. Specifically, the staff used KENO-VI, NITAWL-
1, BONAMI, XSDRN, and ORIGEN. The two-dimensional transport theory code NEWT (Ref. 3)
was used for Boiling Water Reactor (BWR) lattice depletion studies. NEWT uses the method of
characteristics to exactly represent the two-dimensional geometry of the problem. NEWT uses
ORIGEN for depletion. Cross section data were tracked and used on a pin cell basis for the
BWR assessments. The staff developed post processing codes to extract the information from
NEWT and create an input file suitable for use with SCALE. Both the 238 and the 44 group
ENDF/B-V based libraries were used in the project. Refer to Sample Input Deck at the end of
Appendix 7 for a listing of one of the input decks used in this analysis. SCALE has been
extensively validated for these types of assessments. (see References 4, 5, and 6)

Problem Definition

Compression (or expansion) events were analyzed in two ways. First, the assembly was
assumed to crush equally in the x and y directions (horizontal plane). Analyses were
performed with and without the fixed absorber panels without soluble boron and with fuel at the
most reactive point allowed for the configuration. In these cases, the fuel pin pitch was altered
to change the fuel to moderator ratio. These scenarios are intended to simulate the crushing
(or expansion) of a high density configuration when little or no rack deformation is necessary to
apply force to the fuel assembly. The scenarios are also applicable to low density rack
deformation in which the rack structure collapses to the point at which force is applied to the
assemblies. The second type of compression event involved changing the intra-assembly
spacing, but leaving the basic lattice geometry unchanged. These simulations were intended to
simulate compression events in which the force applied to the rack is insufficient to compress
the assembly.

Discussion Of Results

Several observations are common to both Pressurized Water Reactor (PWR) and BWR rack
designs. First of all, poisoned racks should remain subcritical during all compression type
events assuming that the poison sheeting remains in place (in other words, that it compresses
with the rack and does not have some sort of brittle failure). Secondly, criticality cannot be
precluded by design following a compression event for low density, unpoisoned (referring to
both soluble and fixed poisons) storage racks.

PWR Spent Fuel Storage Racks

The analyses and this discussion will differentiate between high and low density storage. High
density storage is defined as racks that rely on both fixed poison sheets and geometry to
control reactivity and low density storage relies solely upon geometry for reactivity control. The
results of the analyses for the high density storage racks are summarized in Figure 1. When
discussing Figure 1 it should be noted that the analyses supporting Figure 1 were performed
without soluble boron and with fuel at the most reactive point allowed for the rack. These
assumptions represent a significant conservatism of at least 20 percent delta-k. Figure 1
demonstrates that even with compression to an optimal geometric configuration, criticality is
prevented by design (for these scenarios we are not trying to maintain a k._ less than 0.95).
The poison sheeting, boral in this case, is sufficient to keep the configuration subcritical.

Appendix 3 A3-6


and the assumption that the leak is self-limiting after a drop in level of 15 feet. These
assumptions may be nonconservative on a plant-specific basis depending on SFP
configuration and commitments for configuration control.

— IDC #10 is credited for the equipment availabilities and relibilities used in the analysis. In
addition, if there are administrative procedures to control the out-of-service duration for
the diesel fire pump, the relatively high unavailability for this pump (0.18) could be
lowered.

+ Initiating event frequencies for loss of cooling, loss of inventory, and loss of offsite power
are based on generic data. In addition, the probability of power recovery is also based on
generic information. Site-specific differences will proportionately affect the risk from these
initiating events.

The various initiating event categories are discussed below. The staffs qualitative tisk insights
on the potential for SFP criticality are discussed in Section 3.6.

3.4 Internal Event Scenarios Leading to Fuel Uncovery

This section describes the events associated with internal event initiators. More details are
given in Appendix 2A.

3.4.1 Loss of Cooling

The loss of cooling initiating event may be caused by the failure of pumps or valves, by piping
failures, by an ineffective heat sink (e.g., loss of heat exchangers), or by a local loss of power
(e.g., electrical connections). Although it may not be directly applicable because of design
differences in decommissioning plants, operational data from NUREG-1275, Volume 12 (Ref. 3),
shows that the frequency of loss of SFP cooling events in which temperature increases more
than 20 °F is on the order of two to three events per 1000 reactor years. The data also shows
that was the loss of cooling lasted less than 1 hour. Only three events exceeded 24 hours: the
longest was 32 hours. In four events the temperature increase exceeded 20 °F, the largest
increase being 50 °F.

The calculated fuel uncovery frequency for this initiating event is 1.4x10° per year. Indications
of a loss of pool cooling available to fuel handlers include control room alarms and indicators,
local temperature measurements, increasing area temperature and humidity, and low pool water
level from boiloff. For a fuel uncovery, the plant fuel handlers must fail to recover the cooling
system (either fail to notice the loss of cooling indications or fail to repair or restore the cooling
system). In addition, the fuel handlers must fail to provide makeup cooling using other onsite
sources (e.g., fire pumps) or offsite sources (e.g., a fire brigade). A long time is available for
these recovery actions. In the case of 1-year-old fuel (i.e., fuel that was in the reactor when it
was shutdown 1 year ago), approximately 195 hours is available for a PWR and 253 hours for a
BWR before the water level drops to within 3 feet of the spent fuel. If the fuel most recently
offloaded is only 2 months out of the reactor, the time available is still long (100-150 hours), and
the likelihood of fuel handler success is still very high. These heatup and boiloff times are about
double those reported by the staff before a correction in the staff's heat load assumptions.

3-12


4. EVALUATION OF THE LOAD PATH

Just because a heavy load is dropped does not mean that it will drop on the SFP wall or on the
pool floor. It may drop at other locations on its path. A load path analysis is plant-specific. In
NUREG-0612 it was estimated that the heavy load was near or over the SFP for between

5 percent and 25 percent of the total path needed to lift, move, and set down the load. It was
further estimated that if the load were dropped from 30 feet or higher (or in some circumstances
from 36 feet and higher depending on the assumptions) when it is over the pool floor, and if a
plant-specific load drop analysis had not been performed,° then damage to the pool floor would
result in loss-of-inventory. In addition we looked at the probability that the load drop occurred
over the pool wall from 8 to 10 inches above the edge of the pool wall. In our analysis we
evaluated the chances the load was raised sufficiently high to fail the pool and evaluated the
likelihood that the drop happened over-a vulnerable portion of the load path. Table A2c-2
presents the results for a heavy load drop on or near the SFP. Based on NUREG-0612, if the
cask were dropped on the SFP floor, the likelihood of a loss-of-inventory given the drop is 1.0.
Based on the evaluation presented in NUREG/CR-5176 (Ref. 5), if the load were dropped on
the SFP wall, the likelihood of a loss-of-inventory given the drop is 0.1.

5. CONCLUSION

Our heavy load drop evaluation is based on the method and fault trees developed in
NUREG-0612. New 1990s Navy data were used to quantify the failure rate of the lifting
equipment. The WIPP human error evaluation was used to quantify the failure to secure the
load. We estimated the mean frequency of a loss-of-inventory from a cask drop onto the pool
floor or onto the pool wall from a single failure-proof system to be 2.0x107 per year for 100 lifts
per year.

However, only some of the plants that will be decommissioning plants in the future currently
have single failure-proof systems. Historically, many facilities have chosen to upgrade their
crane systems to become single failure-proof. However, this is not an NRC requirement. The
guidance in NUREG-0612, phase 2 calls for systems to either be single failure-proof or if they
are non-single failure-proof to perform a load drop analysis. The industry through NEI has
indicated that it is willing to commit to follow the guidance of all phases of NUREG-0612.

For licensees that choose the non-single failure-proof handling system option in NUREG-0612,
we based the mean frequency of a loss-of-inventory event on the method used in NUREG-0612.
in NUREG-0612, an alternate fault tree than that used for the single failure-proof systems was
used to estimate the frequency of exceeding the release guidelines (loss-of-inventory) for a non-
single failure-proof system. We calculated the mean frequency of catastrophic pool failure (for
drops into the pool, or on or near the edge of the pool) for non-single failure-proof systems to be

3 if a load drop analysis were performed, it means that the utility has evaluated the plant
design and construction to pick out the safest path for the movement of the heavy load. In
addition, it means that the path chosen has been evaluated to assure that if the cask were to
drop at any location on the path, it would not catastrophically fail the pool or its support systems.
If it is determined that a portion of the load path would fail if the load were dropped, the as-built
plant must be modified (e.g., by addition of an impact limiter or enhancement of the structural
capacity of that part of the building) to be able to take the load drop or a different safe load path
must be identified.

Appendix 2C. A2C-3


APPENDIX 2D
STRUCTURAL INTEGRITY OF SPENT FUEL POOL STRUCTURES SUBJECT TO AIRCRAFT
CRASHES

1. INTRODUCTION

The mean frequency for significant PWR or BWR spent fuel pool (SFP) damage resulting from a
direct hit from an aircraft was estimated based on the point target model for a 100 x 50-foot pool
to be 4.1x10° per year. The estimated frequency of loss of support systems leading to SFP
uncovery is bounded by other initiators.

2. ANALYSIS

A detailed structural evaluation of how structures will respond to an aircraft crash is beyond the
scope of this effort. The building or facility characteristics were chosen to cover a range typical
of an SFP that is contained in a PWR auxiliary building or a BWR secondary containment
structure. In general, PWR SFPs are located on, or below grade, and BWR SFPs, while
generally elevated about 100 feet above grade, are located inside a secondary containment
structure. The vulnerability of support systems (power supplies, heat exchangers and makeup
water supplies) requires a knowledge of the size and location of these systems at
decommissioning plants, information not readily available. However, we believe this analysis is
adequately broad to provide a reasonable approximation of decommissioning plant vulnerability
to aircraft crashes.

The staff used the generic data provided in DOE-STD-3014-96 (Ref. 1) to assess the likelihood
of an aircraft crash into or near a decommissioned SFP. Aircraft damage can affect the
structural integrity of the SFP or the availability of nearby support systems, such as power
supplies, heat exchangers, and makeup water sources, and may also affect recovery actions.

The frequency of an aircraft crashing into a site, F, was obtained from the four-factor formula in
DOE-STD-3014-96, and is referred to as the effective aircraft target area model:

F= Y Nix . Pi . fi (X, y): Aj Equation A2d-1
ijk
where:
N, estimated annual number of site-specific aircraft operations (no./yr)
Py = aircraft crash rate (per takeoff and landing for near-airport phases) and
per flight for in-flight (nonairport) phase of operation

fin Y) = aircraft crash location probability (per square mile)
Ay = site-specific effective area for the facility of interest, including skid and fly-

in effective areas (square miles)

(index for flight phase): i=1,2, and 3 (takeoff, in-flight, landing)

(index for aircraft category, or subcategory)

(index for flight source): there could be multiple runways and nonairport
operations

i
J
k

Appendix 2D A2D-1


The staff realizes that the volumetric rate of air flow that a fuel bundle receives during a loss of
cooling event significantly influences the heatup of the bundle. To achieve sufficient long-term
air cooling of uncovered spent fuel, two conditions must be met: (1) an air flow path through the
bundles must exist, and (2) sufficient SFP building ventilation flow must be provided. The
presence of more than about 1 foot of water in the SFP, as in a seismically induced SFP failure
or the late states of a boildown sequence, would effectively block the air flow path. Seismically
induced collapse of the SFP building into the SFP could have a similar effect. Loss of building
ventilation would tend to increase fuel heatup rates and maximum fuel temperatures, as
described in Appendix 1A.

Based on engineering judgment, we have partitioned the frequency of each sequence into two
parts: where the bundles in the spent fuel pool area receive two building volumes of air per hour
(high air flow) and where the bundles receives little or no air flow (low air flow). Table 3.2
provides this partition.

3-8


APPENDIX 2A
DETAILED ASSESSMENT OF RISK FROM DECOMMISSIONING PLANT
SPENT FUEL POOLS

1.0 INTRODUCTION

In Reference 1, the NRC performed a preliminary study of spent fuel pool risk at
decommissioning plants to: examine the full scope of potentially risk-significant issues; identify
credible accident scenarios; document the assessment for public review; and to elicit feedback
from all stakeholders regarding analysis assumptions and design and operational features
expected at decommissioning plants. In the February 2000 draft risk assessment, the staff
updated the June 1999 preliminary draft risk assessment to include industry commitments. In
this current analysis, the February 2000 draft was updated based on:

° stakeholder feedback

. additional thermal-hydraulic calculations

This updated PRA addresses the following initiating events:

. loss of SFP cooling

. fire leading to loss of SFP cooling

. loss of offsite power because of plant centered and grid related causes

. loss of offsite power because of severe weather

. non-catastrophic loss of SFP inventory

External events such as earthquakes, aircraft crashes, heavy load drops, and tornado strikes
that could lead to catastrophic pool failure are dealt with elsewhere in this study. The analysis is
based on the following input. The assumed system configuration is typical of the sled-mounted
systems that are used at many current decommissioned plants. Information about existing
decommissioned plants was gathered from project managers (NRC Staff) of decommissioning
plants, and during visits to four sites covering all four major nuclear steam supply system
vendors (General Electric, Westinghouse, Babcock & Wilcox, and Combustion Engineering).
The assumptions made about the operation of the facility are based in part on a set of
commitments made by NEI (Ref. 2), supplemented by an interpretation of how some of those
commitments might be applied.

2.0 SYSTEM DESCRIPTION

Figure 2.1 is a simplified drawing of the system assumed for the development of the model. The

spent fuel pool cooling (SFPC) system is located in the SFP area and consists of motor-driven

Appendix 2A A2A-1


Comment #87: The consequences should be re-evaluated to include an off-site radiological
release from an on-site fire involving radioactive material from a resin container fire; fire in a
waste storage building; and fire in a container vehicle with waste stored in it that could trigger
emergency response mechanisms.

Response: This evaluation is beyond the scope of this study which is focused on spent fuel
pool accident risk. However, before offsite EP at decommissioning plants could be eliminated,
a licensee would need to perform reviews at their facilities to ensure that there are no other
possible accidents that could result in off-site consequences exceeding EPA protective action
guidelines per existing requirements under 10 CFR 30.32 (i) and 10 CFR 30.72.

Comment #88: Decommissioning nuclear power plants should be evaluated for fires in the low
level waste storage (LLW) area. This stakeholder states that large amounts of LLW could be
stored in on-site LLW storage areas if off-site waste disposal sites are lost by a licensee “mid-
stream” during the decommissioning process.

Response: As part of the staff's broad-scope decommissioning regulatory improvement effort,
the staff will ensure that regulations are in place that would reasonably preclude threats to the
public health and safety from accidents that are significantly less severe than a spent fuel pool
zirconium fire but perhaps more probable, such as the LLW fire described above. To address
the specific concern of the public stakeholder, 10 CFR 50.48 requires decommissioning nuclear
power plant licensees to maintain a fire protection program to address fires which could cause
the release or spread of radioactive materials which could result in a radiological hazard. In
addition, nuclear power plants are also subject to the Commission’s regulations for byproduct
materials under 10 CFR Part 30. Specifically, 10 CFR 30.32(i) would require a licensee to
maintain an appropriate EP program for radioactive materials stored on-site in quantities in
excess of those specified in 10 CFR 30.72, “Schedule C - Quantities of Radioactive Material
Requiring Consideration of the Need for an Emergency Plan for Responding to a Release.”

Appendix 6 A6-29


If the loss of inventory is discovered through walk-downs, it is assumed that there is not enough
time available to isolate the leak in time to provide for SFP makeup system success, and this
event does not appear on the failure branch of event CRA.

4.5.5.2 Relevant Assumptions

. System drawings are kept up to date and training plans are revised as needed to reflect
changes in plant configuration.

¢ With an assumed leak rate of 60 gpm, the operator has in excess of 4 hrs to isolate the
leak and provide make-up.

. There are procedures to guide the operators in how to deal with loss of inventory, and the
operators are trained in their use (NEI commitment no. 2).

. SFP operations that have the potential to rapidly drain the pool will be under strict
administrative controls (NEI commitment no. 9). This increases the likelihood of the
operators successfully terminating a leak should one occur.

4.5.5.3 Quantification
Human Error Probabilities

Two human failure events are included in the functional fault tree LOI-OIS-U, one for failure to
start the SFP makeup pump, HEP-MKUP-START-E, and one for failure to successfully isolate
the leak, HEP-LEAK-ISO.

SPAR HRA worksheets were used to quantify each of these errors. For HEP-MKUP-START-E,
it was assumed that the operator is experiencing a high stress level, he is highly trained, the
equipment associated with the task is well labeled and matched to a quality procedure, and the
crew has effective interactions in a quality facility.

For HEP-LEAK-ISO, it was assumed that the operators would be experiencing a high level of
stress, the task is highly complex because of the fact that it is necessary to identify the source of
the leak and it may be difficult to isolate, the operators are highly trained, have all the equipment
available, and all components are well labeled and correspond to a procedure, and the crew has
effective interactions in a quality facility.

Hardware Failure Probabilities
Unavailability of an SFP makeup system, SFP-REGMKUP-F, was assigned a value of 5.0E-2

from INEL-96/0334. It is assumed that the SFP makeup system is maintained since it is
required often to provide make-up.

Appendix 2A A2A-58


APPENDIX 4C
POOL PERFORMANCE GUIDELINE

1. INTRODUCTION

The Pool Performance Guideline (PPG) provides a threshold for controlling the risk from a
decommissioning plant spent fuel pool (SFP). By maintaining the frequency of events leading
to uncovery of the spent fuel at a value less than the recommended PPG value of 1x10° per
year, zirconium fires will remain highly unlikely, the risk will continue to meet the Commission’s
Quantitative Health Objectives [1], and changes to the plant licensing basis that result in very
small increases in LERF may be permitted consistent with the logic in Regulatory Guide 1.174
[2]. The purpose of this appendix is to present the rationale for the PPG, and to illustrate how
conformance with the recommended PPG will assure that SFP risk in decommissioning plants
will continue to meet the Commission’s quantitative health objectives (QHOs).

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis," contains general
guidance for application of PRA insights to the regulation of nuclear reactors. The same
concepts can also be applied in the regulation of SFPs. The guidelines in RG 1.174 pertain to
the frequency of core damage accidents (CDF) and large early releases (LERF). For both CDF
and LERF, RG 1.174 contains guidance on acceptable values for the changes that can be
allowed as a function of the baseline frequencies. For example, if the baseline CDF for a plant
is below 1x10“ per year, plant changes can be approved that increase CDF by up to 1x10 per
year. If the baseline LERF is less than 1x10° per year, plant changes can be approved that
increase LERF by up to 1x10* per year.

For decommissioning plants, the risk is primarily because of the possibility of a zirconium fire
associated with the spent fuel cladding. The consequences of such an event do not equate
directly to either a core damage accident or a large early release as modeled for an operating
reactor. Zirconium fires in SFPs have the potential for significant long-term consequences
because: there may be multiple cores involved; the relevant clad/fuel degradation mechanisms
could lead to increased releases of certain isotopes (e.g., short-lived isotopes such as iodine
will have decayed, but the release of longer-lived isotopes such as ruthenium could be .
increased because of air-fuel reactions); and there is no containment surrounding the SFP to
mitigate the consequences. On the other hand, they are different from a large early release
because the postulated accidents progress more slowly, allowing time for protective actions to
be taken to significantly reduce early fatalities (and to a lesser extent latent fatalities). In effect,
an SFP fire would result in a "large" release, but this release would not generally be considered
"early" because of the significant time delay before fission products are released.

Even though the event progresses more slowly than an operating reactor large early release
event and the isotopic make-up is somewhat different, the consequence calculations performed
by the staff (reported in Appendix 4 and 4A) show that SFP fires could have significant health
effects on par with those for a severe reactor accident. These calculations considered the
effects of different source terms, evacuation assumptions, and plume-related parameters on
offsite consequences. Since an SFP fire scenario would involve a direct release to the
environment with significant consequences, the staff has decided that the RG 1.174 LERF
baseline guideline of 1x10° per year (the value of baseline risk above which the staff will only
consider very small increases in risk) provides an appropriate threshold for controlling the risk
from a decommissioning plant SFP, and has established 1x10° per year as the recommended
PPG for this purpose. Maintaining the frequency of events leading to uncovery of the spent fuel
at a value less than the PPG, will assure that zirconium fires remain highly unlikely and that the
tisk in a decommissioning plant will continue to meet the Commission’s QHOs, as discussed
below. Conformance with the PPG is also essential if the staff is to permit changes to the

Appendix 4C A4C-1


. There are procedures to guide the operators in how to deal with loss of inventory, and the
operators are trained in their use. The guidance on when to begin addition of water from
alternate sources is clear and related to a clearly identified condition, such as pool level or
onset of boiling (NEI commitment no. 2).

4.5.7.3 Quantification
Human Error Probabilities

Each fault tree includes three human failure events. In the case of a functional fault tree
LOI-OMK-SMLK, a basic event HEP-RECG-FWSTART represents the failure of the operator to
recognize the need to initiate firewater as an inventory makeup system; a basic event
HEP-FW-START represents failure to start either the electric or diesel firewater pump; and a
basic event HEP-FW-REP-NODSM represents the failure of the repair crew to repair a firewater
pump.

For functional fault tree LOI-OMK-SMLK-L, the basic event HEP-RECG-FWSTART is replaced
by HEP-RECG-FWSTART-L. This event requires that the operators recognize that the
deteriorating conditions in the SFP are because of an inventory loss. The cues will include pool
heat up because of the loss of SFP cooling which should be alarmed in the control room, as well
as other physical indications such as increasing temperature and humidity, and a significant loss
of level. Because of the nature of the sequence, the failure to recognize the need for action will
be modeled by assuming a low dependence between this event and the prior failures.

For functional fault tree LOI-OMK-LGLK, a basic event HEP-RECG-FW-LOI represents the
failure of the operator to recognize the need to initiate firewater as an inventory makeup system;
a basic event HEP-FW-START-LOI represents failure to start either the electric or diesel
firewater pump; and a basic event HEP-FW-REP-NODLG represents the failure of the repair
crew to repair a firewater pump.

SPAR HRA worksheets were also used to quantify the HEPs.

HEP-FW-START represents failure to start either the electric or diesel firewater pump
(depending upon availability), given that the decision to start a firewater pump was made. No
difficult valve alignment is required, but the operator may have to run hoses to designated valve
stations, therefore, expansive time is assumed, with all other PSFs being the same as the other
HEPs below.

For HEP-RECG-FWSTART it was assumed that extensive time is available to the operators for
diagnosis, that the operators are under high stress, are highly trained, have a diagnostic
procedure, have good instrumentation in the form of alarms, and are part of a crew that interacts
well in a quality facility.

For HEP-RECG-FW-LOI it was assumed that extra time (>60 minutes) is available to the
operators for diagnosis, that the operators are under high stress, are highly trained, have a
diagnostic procedure, have good instrumentation in the form of alarms, and are part of a crew
that interacts well in a quality facility. r

Appendix 2A A2A-61


Table 4.5 Basic Event Summary for the Loss of Inventory Event Tree

failure

. age Basic Event
Basic Event Name Description Probability
lE-LOI Loss of inventory initiating event 1.0E-3

Operators fail to respond to a signal indication
HEP-DIAG-LGLK in the contro! room (large leak) 4.0E-4

_ Operators fail to respond to a signal indication

HEP-DIAG-ALARM in the control room 3.0E-4

Operators fail to observe the LOi/loss of
HEP-WLKDWN-LOI cooling in walk-downs, given failure to prevent 5.0E-3

loss of SFP cooling

Operators fail to observe the LOI/loss of
HEP-WLKDWN-LOML. cooling in walk-downs (independent case) 1.0E-5

Operators fail to observe the LO! event walk-
HEP-WLKDWN-DEPEN downs (dependent case) 5.0E-2

Operators fail to diagnose need to start the
HEP-RECG-FW-LOI firewater system 2.0E-4
HEP-RECG-FWSTART Operators fail to diagnose need to start the 20E-5

firewater system

Operators fail to diagnose need to start the
HEP-RECG-FWSTART-L | firewater system given he failed to prevent 5.0E-2

loss of SFP cooling
HEP-LEAK-ISO Operators fail to isolate leak 1.3E-3
HEP-FW-START-LOI Fails to start firewater pumps 1.3E-3

EW. Operators fail to start firewater pump and _
HEP-FW-START provide alignment 1.0E-5
HEP-FW-REP-NODLG Fails to repair firewater pump (20 hrs) 9.0E-2
HEP-FW-REP-NODSM Fails to repair firewater pump (49 hrs) 7.5E-3
HEP-INV-OFFST-LK Operators fail to recover via offsite sources 5.0E-2

INVA Operators fail to provide alternate sources of _
HEP-INV-OFFSITE cooling from offsite 5.0E-2
FP-2PUMPS-FTF Failure of firewater pump system 6.7E-4
LOI-LGLK Loss exceeds normal make-up 6.0E-2
HEP-MKUP-START Operators fail to start make-up(small leak) 2.5E-6
HEP-MKUP-START-E Roa fail to start make-up(Early 25E-4
HEP-MKUP-START-L Operators fail to start make-up(Late Respond) 1.0
SFP-REGMKUP-F _| Regular SFP make-up system fails 5.0E-2
SPC-LVL-LOP Electrical faults leading to alarm channel 2.0E-3

Appendix 2A

A2A-65



curve, the mean value of the pool failure frequency is about 2x10” per year. In considering
these two different sets of hazard estimates, the NRC has found that both sets are reasonable
and equally valid.

By passing the checklist, the SFP will be assured a high confidence with low probability of failure
(HCLPF)" of at least 1.2 g peak spectral acceleration. The performance of the seismic checklist
is identified by the staff as SDA #6.

SDA #6 Each decommissioning plant will successfully complete the seismic checklist provided
in Appendix 2B to this study. If the checklist cannot be successfully completed, the
decommissioning plant will perform a plant specific seismic risk assessment of the
SFP and demonstrate that SFP seismically induced structural failure and rapid loss of
inventory is less than the generic bounding estimates provided in this study (<1x10%
per year including non-seismic events).

For many sites (particularly PWRs because their SFPs are closer to ground level or embedded
and the motion is therefore less amplified), the plant-specific risk may be considerably lower.
There are only two plant-specific SFP fragility analyses of which the staff is aware, and these
were used in the analyses performed to help confirm the generic seismic capacities assumed for
SFPs.

“The HCLPF value is defined as the peak seismic acceleration at which there is
95 percent confidence that less than 5 percent of the time the structure, system, or component
will fail.

3-20


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Table A4-2 Half-lives of MACCS Radionuclides Whose Inventories Were Not in

NUREG/CR-4982
Radionuclide | Half-Life
(days)

Kr-85m 19
Kr-87 05
Kr-88 12
Sr-91 40
Sr-92 11
Y-92 15
Y-93 42
Zr-97 70
Ru-105 19
Rh-105 1.48
Sb-129 18
Te-131m 1.25
1-133 87
1-134 .04
1-135 27
Xe-135 -38
Ba-139 .06
La-141 16
La-142 07
Ce-143 1.38

Release Fractions

NUREG/CR-4982 also provided the fission product release fractions assumed for a severe
SFP accident. These fission product release fractions are shown in Table A4-3. NUREG/CR-
6451 provided an updated estimate of fission product release fractions. The release fractions
in NUREG/CR-6451 (also shown in Table A4-3) are the same as those in NUREG/CR-4982,
with the exception of lanthanum and cerium. NUREG/CR-6451 stated that the release fraction
of lanthanum and cerium should be increased from 1x10* in NUREG/CR-4982 to 6x10%,

Appendix 4 A4-5


Table A2c-1 Summary of the 1996-1999 Navy Crane Data

INon-rigging| Rigging Tota!
[summary by Incident Type (fraction of events) ID Fraction | Fraction | Fraction
[Crane collision cc 0.17 0.00 0.17
Damaged crane oc 0.20 0.08 0.27
[Damaged load DL 0.02 0.03 0.05
Dropped toad op 0.03 0.06 0.09
Load collision uc 0.11 0.03 0.14
[Other 00 0.02 0.00. 0.02
Overload OL 0.08 0.05 0.12
[Personnel injury PI 0.03 0.05 0.08
[Shock Sk 0.00 0.02 0.02
[Two-blocking 1B 0.05 0.00 0.05
[Unidentified Tr) 0.02 0.00 0.02
otals 0.70 0.30 1.00
[Summary by Incident Cause (fraction of total events) [10 | Fraction | I
[improper operation 10 0.38
[Procedures Proc | _0.20
[Equipment failure EQ 0.05
improper rigging” IR 0.30
[Others - OTHER| 0.08
[Totals 1.00
Fault Tree ID® [Application of new Navy data to heavy load drop evaluation | Fraction | [_NUREG-0612 Fraction
Fi {OL + 0.5*(DL+LC) 0.14 0.05
F2 ICC + DC + 0.5(DL+LC) + DD + OO + Pl + SK+ UD + 0.3°IR 0.61 0.53
F3 ie 0.05 0.35,
F4 [Assume next incident (0.01) (1/44)
F5 Rigging 0.7°IR 0.21 0.07
[rotals 1.00 1.00

Notes:

Based on database description, 30 percent or “improper rigging” by incident cause were rigging failures
during crane movement, and 70 percent of “improper rigging” by incident cause were rigging errors.

F1 - Load hangup resulting from operator error (assume 50 percent of “damaged load” and “load collision” lead to hangup)
F2 - Failure of component with a backup component (assume 50 percent of “damaged load” and “load collision” lead to

component failure)

F3 - Two-blocking event
F4 - Failure of component without a backup
F§ - Failure from improper rigging

A2C-6



10.

Procedures will be in place to establish communication between onsite and offsite
organizations during severe weather and seismic events.

An offsite resource plan will be developed which will include access to portable pumps and
emergency power to supplement onsite resources. The plan would principally identify
organizations or suppliers where offsite resources could be obtained in a timely manner.

SFP instrumentation will include readouts and alarms in the control room (or where
personnel are stationed) for SFP temperature, water level, and area radiation levels.

SFP boundary seals that could cause leakage leading to fuel uncovery in the event of seal
failure shall be self limiting to leakage or otherwise engineered so that drainage cannot
occur.

Procedures or administrative controls to reduce the likelihood of rapid draindown events
will include (1) prohibitions on the use of pumps that lack adequate siphon protection; or
(2) controls for pump suction and discharge points. The functionality of anti-siphon
devices will be periodically verified.

An onsite restoration plan will be in place to provide for repair of the SFP cooling systems
or to provide access for makeup water to the SFP. The plan will provide for remote
alignment of the makeup source to the SFP without requiring entry to the refuel floor.

Procedures will be in place to control SFP operations that have the potential to rapidly
decrease SFP inventory. These administrative controls may require additional operations
or administrative limitations such as restrictions on heavy load movements.

Routine testing of the alternative fuel pool makeup system components will be performed
and administrative controls for equipment out of service will be implemented to provide
added assurance that the components would be available if needed.

Since the commitments are stated at a relatively high level, additional assumptions have been
made as detailed below.

It is assumed that the operators (through procedures and training) are aware of the
available backup sources that can be used to replenish the SFP inventory (i.e., the fire
protection pumps, or offsite sources such as from fire engines). Arrangements have been
made in advance with fire stations including what is required from the fire department
including equipment and tasks.

The site has two operable firewater pumps, one diesel-driven and one electrically driven
from offsite power.

The makeup capability (with respect to volumetric flow) is assumed as follows:

Makeup pump: 20-30 gpm
Firewater pump: 100 - 200 gpm

Appendix 2A . A2A-11


em-242 100.3017E-06 300.00 end

cm-244 10 0.6846E-06 300.00 end

i-135 10 0.2543E-07 300.00 end

‘Zire

cr 207.5891E-5 300.0 end

fe 2 0 1.4838E-4 300.0 end

ar 2.0 4.2982E-2 300.0 end

‘Water w/ 2000 ppm boron

h2o 3 0.99 300.0 end

'b-10  302.2061E-5 300.0 end

‘SS structural material

ss304 40.99 300.0 end

"Boral (model as b4c-al using areal density of b-10 @ -- g/em*2 and 0.18 atom percent b-10 in
nat. b)

‘Excluded Proprietary Information

end comp

‘squarepitch card excluded - Proprietary Information
more data

dab=999

end more

read param

gen=103 npg=3000 xs1=yes pki=yes gas=yes fix=yes fdn=yes far=yes nb8=999
end param

read geom

‘geom cards excluded - Proprietary Information

end geom

read array

ara=1 nux=15 nuy=15 nuz=" fill
1144141414144 41414414 1 41
114141414144 4 4144 14 4 4 «41
1121412143411 2414414 2 41~«421
1141414141411 2434434414444 «424
1114142141414 3% 24<4«4 4
142414141 141% 1414414 2414
1414141414144 1% 414414414 414 «414 4
414124141 214412414441
4114414144144 4 41414414414 «421
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1141421414144 1244414421
1141414141414 2441444 4414421
1121121414442 14414 241421
144414 14414 414 44 4 414 4 «4 «41
14141414 1441441441414 414 414 «4

end fill

end array

read bounds all=mirror end bounds
read mixt sct=2 eps=1.e-01 end mixt
read plot

scr=yes

Appendix 3 A3-12


FALS TO RECOGNIZE
NEED TO START FW
PUMPS

HERE

A-21

FAILURE TORECOVER
VA OFFSITE SOURCES.



Comment #45: Has the NRC considered the events with the “second” worst off-site
consequences at decommissioning plants? For example, in another country which has nuclear
power plants, a fire in the bitumen storage (waste handling area) was found to have the second
worst, although limited, off-site consequences.

Response: This study evaluated a spectrum of potentially severe spent fuel pool accidents.
However, before offsite EP at a decommissioning plant could be eliminated, a licensee would
need to perform reviews of their facilities to ensure that there are no other possible accidents
that could result in off-site consequences exceeding the EPA protective action guidelines per
existing requirements under 10 CFR 30.32(i) and 10 CFR 30.72.

Appendix 6 AG-14


Appendix 3

Sample Input Deck Listing and
Tables and Figures

A3-10


Assessment of the Potential for Criticality in
Decommissioned Spent Fuel Pools

Tony P. Ulses
Reactor Systems Branch
Division of Systems Safety and Analysis

Appendix 3. A3-4


APPENDIX 4
CONSEQUENCE ASSESSMENT FROM ZIRCONIUM FIRE

Spent fuel pool (SFP) accidents involving a sustained loss of coolant have the potential for
leading to significant fuel heat up and resultant release of fission products to the environment.
Such an accident would involve decay heat raising the fuel temperature to the point of
exothermic cladding oxidation, which would cause additional temperature escalation to the
point of fission product release. However, because fuel in an SFP has a lower decay power
than fuel in the reactor vessel of an operating reactor, it will take much longer for the fuel in the
SFP to heat up to the point of releasing radionuclides than in some reactor accidents.

Earlier analyses in NUREG/CR-4982 (Ref. 1) and NUREG/CR-6451 (Ref. 2) have assessed
the frequency and consequences of SFP accidents. These analyses included a limited
evaluation of offsite consequences of a severe SFP accident. NUREG/CR-4982 results
included consequence estimates for the societal dose for accidents occurring 30 days and 90
days after the last discharge of spent fuel into the SFP. NUREG/CR-6451 results included
consequence estimates for societal dose, prompt fatalities, and cancer fatalities for accidents
occurring 12 days after the last discharge of spent fuel. The work described in this appendix
extends the earlier analyses by calculating offsite consequences for a severe SFP accident
occurring up to one year after discharge of the last load of spent fuel, and supplements that
earlier analysis with additional sensitivity studies, including varying evacuation assumptions as
well as other modeling assumptions. The primary objective of this analysis was to assess the
effect of extended storage in an SFP, and the resulting radioactive decay, on offsite
consequences. However, as part of this work, the sensitivity to a variety of other parameters
was also evaluated.

The current analysis used the MACCS code (Ref. 3)(version 2) to estimate offsite
consequences for a severe SFP accident. Major input parameters for MACCS include
radionuclide inventories, radionuclide release fractions, evacuation and relocation criteria, and
population density. The specification of values for these input parameters for a severe SFP
accident is discussed below.

Radionuclide Inventories

As discussed above, the current analysis was undertaken to assess the magnitude of the
decrease in offsite consequences that could result from up to a year of decay in the SFP. To
perform this work, it was necessary to have radionuclide inventories in the SFP for a
decommissioned reactor at times up to 1 year after final shutdown. The inventories in the
NUREG/CR-6451 analysis have not been retrievable, so those inventories could not be used.
NUREG/CR-4982 contains SFP inventories for two operating reactors, a BWR (Millstone 1)
and a PWR (Ginna). Since the staff had radionuclide inventory data for a small BWR (Millstone
1), the staff adjusted the radionuclide inventory of Millstone 1 to represent a large BWR with a
thermal power of 3441 megawatts. These SFP inventories for Millstone 1 are given in Table
4.1 of NUREG/CR-4982 and are reproduced in Table.A4-1 below. Two adjustments were then
made to the Table A4-1 inventories. The first adjustment was to multiply the inventories by a
factor of 1.7, because the thermal power of the large BWR is 1.7 times higher than that of
Millstone 1. The second adjustment, described in the next two paragraphs, was needed

Appendix 4 A4-1


Loss of Offsite Power
Cy, = 0.10g PGA
Cio, = 0.18g PGA (2)
Cso, = 0.35g PGA

Loss of Even Temporary Safe Usability of Well Designed Buildings and Bridges
Ci, = 0.20g PGA

Cio, = 0.35g PGA (3)
Coo, = 0.75g PGA

Thus, for a 0.5 PGA scenario ground motion, I would expect less than about a 1%
chance of the spent fuel pool failing to hold water, about a 70 to 75% chance that offsite
power to the station is lost, and about 20 to 25% of the well designed surrounding
buildings (housing communication systems) and bridges being unsafe to use even
temporarily. By "well designed", I mean the building or bridge has some form of lateral
load carrying system, but does not have nuclear plant or California levels of seismic
design. Many CEUS buildings and bridges will have lesser seismic capacity than does
this "well-designed" category, and a few might be better. Therefore, over the entire
population of nearby buildings and bridges, I would expect more than 20 to 25% would
be unsafe for even temporary use.

For a 0.75g PGA scenario ground motion, I would expect less than about a 10%
chance of the spent fuel pool failing, about a 90% chance that offsite power is lost, and
more than about 50% of the CEUS buildings and bridges being unsafe for even temporary
use. At this ground motion level which is within the region of ground motions that
dominate the estimated seismic risk of spent fuel pool failures, sufficient power, buildings
housing communication systems and emergency services, and bridges will be out-of-
service that emergency responses will most likely have to be ad-hoc. Specifically, for
ground motion levels that correspond to spent fuel pool failure, within at least 10 miles of
the plant I would expect power to have been lost and more than about 50% of the CEUS
bridges and buildings (including those housing communication systems and emergency
response equipment) being unsafe for even temporary use.

4. Reference
1. Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two

Representative Nuclear Power Plants, NUREG/CR-5176, Prepared for Nuclear
Regulatory Commission, January 1989

A2B-25 -


4.2.6.3 Quantification

Human Error Probabilities

The event HEP-INV-OFFSITE represents failure to recognize that it is necessary to take the
extreme measure of using offsite sources, given that even though there has been ample time up
to this point to attempt recovery of the firewater pump, it has not been successful. This top
event should include failures of both the diagnosis of the need to provide inventory from offsite
sources, and of the action itself. The availability of offsite resources is assumed not to be
limiting on the assumption of an expansive preparation time. However, rather than use a
calculated HEP directly, a low level of dependence to account for the possible detrimental
effects of the failure to complete prior tasks successfully.

4.2.6.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-INV-OFFSITE §.0E-2

4.2.7 Summary
Table 4.2 presents a summary of basic event probabilities used in the event tree quantification.

As in the case of the loss of cooling event, the frequency of fuel uncovery, based on the
assumptions made in the analysis, is very low. The assumptions that support this low value
include: careful and thorough adherence to NEI commitments 2, 5, 8 and 10; walk-downs are
performed on a regular, (once per shift) (important to compensate for potential failures to the
instrumentation monitoring the status of the pool); procedures and/or training are explicit in
giving guidance on the capability of the fuel pool makeup system, and when it becomes
essential to supplement with alternate higher volume sources; procedures and training are
sufficiently clear in giving guidance on early preparation for using the alternate makeup sources.

Appendix 2A A2A-32


FAILURE TOPROVIDE
MAKEUP USING PW
SYSTEM
|
f t 1
FALS TODIAGNOSE FAILURE TO MAEUP OPERATOR FAILS TO
NEEDTO START FW INVENTORY BY USE OF ESTABLISH ALTERNATE
FIRE PROTECTION PUNP ‘COOLING - EARLY
ve. ART 3 Hep PWGraRr
fo 1
ECT FALS TOREPAR
DESL ADB ERC DIESEL ORELECTRIC
& PONE - DePENDENCY
FART AND RUN, ey
FR2PUNPSFTF HEP-FWEREP-DEPEN
n
FAILURE TOPROVIDE
MAKEUP USING PW
‘SYSTEM
1
r t 1
DIAGNOSE FAILURE TO MAKEUP OPERATOR FALS TO
NEE 1S Saki INVENTORY BY USE OF ESTABLISH ALTERNATE
FIRE PROTECTION PUNP COGLNG - EARLY
HEP RECGPWSTARTL : Hep PWSTART
£ 1
ELECTR FALS TOREPAR
ee eae DIESEL OR ELECTRIC
START AND RUN, PUMP DEPENDENCY
FRORNPSFTF HEP-FWREP-DEPEN

A-4


APPENDIX 4A

Memo to Gary M. Holahan from Farouk Eltawila dated August 25, 2000, re:
RISK-INFORMED REQUIREMENTS FOR DECOMMISSIONING

Appendix 4A AGA-1


Comments #41 and 42: (A) The staff's spent fuel pool risk study only considered accident
scenarios that could lead to a spent fuel zirconium fire and asked what other design basis
accidents are considered for decommissioning nuclear power plants beyond those addressed
in the study. (B) What design basis accidents do we need to consider?

Response: There are typically no new or unique conditions associated with decommissioning
that result in the creation or possibility of a different type of accident not previously bounded by
the design basis accidents considered for the plant while it was operating. When a licensee
updates its Final Safety Analysis Report for decommissioning, a suite of accidents are
considered that have a reasonable potential to adversely impact public health and safety. The
off-site consequences of these accidents are generally very small and should not require off-
site emergency response. Examples of the types of accidents that are considered by the
licensees include:

Materials handling event (non-fuel)
Radioactive liquid waste releases
Accidents from handling spent resin
Fire

Explosions

External events

Transportation accidents

Fuel handling accident

eo eee wee

In addition to plant specific assessments of the postulated accidents, the staff has performed
some generic evaluations. Consideration of environmental impacts of such events has been
provided in NUREG-0586, “Final Generic Environmental Impact Statement on
Decommissioning of Nuclear Facilities.”

Comment #43: A commenter stated that Industry Decommissioning Commitment #5 should be
revised to require direct measurement of SFP temperature and water level.

Response: The staff agrees; SDA #3 calls for direct measurement of SFP temperature and
water level.

Comment #44: Dr. Hanauer was quoted in a 1975 memo to say, “you can make probabilistic
numbers prove anything, by which | mean that probabilistic numbers prove nothing.” Ifa
respected technical advisor has expressed doubts about the NRC’s use of probabilistic
numbers, how is the NRC going to use probabilities convincingly to protect health and safety?
A commenter stated that, “this is an invalid way of measuring safety, and should not be used.
Each day these reactors stay opened you are poisoning the environment. This is
unacceptable.”

Response: in the two and a half decades since this statement, there have been significant
advances in risk assessment methodologies. In that time frame, the NRC has also gained a
great deal of experience in applying these methodologies to the regulatory arena, which has
led to improved safety. The NRC has determined that PRA is an acceptable technology and
uses it in a manner that complements a deterministic approach and supports the traditional
defense-in-depth philosophy.

Appendix 6 A6-13


Comment #102: Untrained NRC public representatives frequently misinform the public,
particularly about the opportunities for a hearing on reactor decommissioning.

Response: The NRC endeavors to train all NRC employees for their specific work
assignments. In the event that misinformation is inadvertently communicated by an individual
staff member, the NRC staff upon identifying the misinformation provides the correct
information in the most expedient manner.

Comment #103: A commenter cited several specific examples of interactions with NRC staff
that he felt demonstrated improper or inaccurate information provided by NRC staff members.

Response: In the course of oral communication with the public in an open and unrestrained
fashion, errors, misspoken words, and misunderstandings may occur by the individuals from
the public and the NRC staff. The NRC endeavors to minimize these miscommunications from
our staff, but should they occur, NRC staff will act to correct them by the most expedient means
available.

Comment #104: The time delays experienced by licensees who must submit individual heatup
analyses and applications for exemption from NRC regulations could be mitigated by
preparation of such documentation well in advance of decommissioning.

Response: It is true that decommissioning licensees who have planned reactor shutdown
schedules far in advance would be able to submit exemption requests and conduct supporting
thermal-hydraulic analyses in advance of reactor shutdown so that lengthy regulatory delays
could be minimized. However, plants that shut down unexpectedly would not be able to submit
such analyses in advance. The NRC believes that it should promulgate new decommissioning
regulations that ensure public health and safety, reduce unnecessary regulatory burden and
increase the efficiency and effectiveness of operations for both licensees and the NRC.

Comment #105: In a letter to the NRC, a commenter stated that the NRC staff owes its
stakeholders the courtesy of addressing their concerns, particularly when comments are
solicited by the NRC staff. Otherwise, the NRC staff must stop actively soliciting public
comment when it has no intention of considering.

Response: At the July 15-16, 1999 public workshop on decommissioning spent fuel pool risk,
the public stakeholder raised a concer that the NRC evaluate potential hazards that
decommissioning accidents could impose upon plant workers. When the NRC issued its final
draft report, the stakeholder’s issue was not specifically addressed in the comment evaluation
section. However, the NRC had received an industry decommissioning commitment that
licensees would provide a remote method of adding water to spent fuel pools that would reduce
potential risk to plant workers and which resulted from the issue the stakeholder had raised.
The NRC seriously considers public comments received on all issues within its jurisdiction. In
this case, the staff regrets the appearance that a public comment had been ignored. In order to
ensure that proper consideration was given to all stakeholder comments, the NRC staff
reviewed written comments received and examined transcripts of public meetings to ensure
that all issues had been addressed.

Comment #106: A commenter requested on April 10, 2000, that the comment period on the

Appendix 6 A6-34


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3-44


agencies such as fire brigades) in parallel with a primary response such as carrying out
repairs or lining up an onsite alternate system.

° clearly and unambiguously written, with an understanding of a variety of different scenarios
and their timing.

In addition:

. training for plant staff to provide an awareness of the time scales of heat up to boiling and
fuel uncovery as a function of the age of the fuel would enhance the likelihood of
successful response.

3.2.3.3 Response Implementation

Successful implementation of planned responses may be influenced by several factors, for
example:

. accessibility/availability of equipment

. staffing levels that are adequate for conducting each task and any parallel contingency
plans, or plans to bring in additional staff

. training

° timely feedback on corrective action

3.2.4 Quantification Method

Three HRA quantification methods were applied, and each is briefly described below.

. The Technique for Human Error Prediction (THERP, Ref. 10). This method was used to
quantify the initial recognition of the problem. Specifically, the annunciator response
model (Table 20-23) was used for response to alarms. The THERP approach was also
used to assess the likelihood of failure to detect a deviant condition during a walk-down,
and also the failure to respond to a fire. While this method was developed over 20 years
ago, it is still regarded as an appropriate method for the types of HEPs for which it is being
used in this analysis.

. The Exponential Repair Model (while not strictly a human reliability model) was applied to
calculate the probability of failure associated with the repair of systems and components in
this analysis. This method is described in the main body of the study. In cases where
dependency exists with prior repair tasks, the dependency model used in THERP was
used to assess the impact of that dependency.

. The Simplified Plant Analysis Risk Human Error Analysis Method (SPAR HRA, Ref. 11)
was employed for all other HEPs. This model was chosen because it includes an
appropriate level of detail in terms of performance shaping factors and error modes

Appendix 2A A2A-8


4.2 Implications for Regulatory Requirements for Emergency Preparedness, Security, and
Insurance

The industry and other stakeholders have expressed interest in knowing the relevance of the
results of this study to decisions on specific regulatory requirements. These decisions could be
made in response to plant-specific exemption requests or as part of the integrated rulemaking
for decommissioning plants. Such decisions can be facilitated by a risk-informed examination of
both the deterministic and probabilistic aspects of decommissioning. Three examples of such
regulatory decisions are presented in this section: regulatory requirements for emergency
preparedness, security, and insurance.

4.2.1 Emergency Preparedness

The requirements for emergency preparedness are contained in 10 CFR 50.47 (Ref. 5) and
Appendix E to 10 CFR Part 50 (Ref. 6). Further guidance on the basis for EP requirements is
contained in NUREG-0396 (Ref. 7) and NUREG-0654/FEMA-REP-1 (Ref. 8). The task force of
NRC and Environmental Protection Agency (EPA) representatives formed to address the
planning bases for emergency preparedness concluded that the overall objective of EP is to
provide dose savings (and in some cases immediate life saving) for a spectrum of accidents that
could produce offsite doses in excess of predetermined protective action guides (PAGs).

In the past, the NRC staff has typically granted exemptions from offsite emergency planning
requirements for decommissioning plants that could demonstrate that they were beyond the
period in which a zirconium fire could occur. The rationale for those decisions was that, in the
absence of a zirconium fire, there were no decommissioning plant scenarios for which the
consequences justify the imposition of an offsite EP requirement. The results of this technical
study confirm that the frequency of events leading to SFP fires is very low (ranging from about
4x107 at sites where seismic events are a minimal contributor to less than 1x10° per year at
sites where seismic events dominate SFP risk and no plant-specific seismic analyses need to be
performed), and that the subset of events in which EP can produce significant dose savings is
even smaller (about 2x10” per year). However, the staff concludes that the possibility an SFP
accident will lead to a large fission product release cannot be ruled out even many years after
final shutdown, since several SFP accidents could involve either blockage of the air cooling path
(e.g., due to partial draining of the SFP) or inadequate air circulation within the SFP building,
resulting in near-adiabatic heatup of the spent fuel. The impact of this new information on
previously granted exemptions is being evaluated by the staff. Large seismic events that fail the
SFP are the dominant contributor to these failure modes. Emergency planning would be of
marginal benefit in reducing the risk of such events due to its impairment by offsite damage.
The next largest contributor, cask drop accidents, is about an order of magnitude lower in
frequency. In the first few years following final shutdown (when time to fission product release is
less than about 10 hours), EP could provide some dose savings, but does not substantially
impact risk due to the low frequency of these events. Finally, although large releases from the
SFP would remain possible for these failure modes, the time available before release would be
in excess of 24 hours 5 years after final shutdown and sufficient to support implementation of
protective measures on an ad hoc basis.

4-13


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Appendix 4C.


SDA #3 Control room instrumentation that monitors SFP temperature and water level directly
measures the parameters involved. Level instrumentation provides alarms for calling
in offsite resources and for declaring a general emergency.

3.4.2 Loss of Coolant Inventory

This initiator includes loss of coolant inventory resulting from configuration control errors,
siphoning, piping failures, and gate and seal failures. Operational data in NUREG-1275,
Volume 12, shows that the frequency of loss of inventory events in which a level decrease of
more than 1 foot occurred is less than one event per 100 reactor years. Most of these events
are as a result of fuel handler errors and are recoverable. Many of the events are not applicable
in a decommissioning facility.

NUREG-1275 shows that, except for one event that lasted 72 hours, no events lasted more than
24 hours. Eight events resulted in a level decrease of between 1 and 5 feet, and another two
events resulted in an inventory loss of between 5 and 10 feet.

Using the information from NUREG-1275, it can be estimated that 6 percent of the loss of
inventory events will be large enough and/or long enough to require that isolating the loss if the
only system available for makeup is the SFP makeup system. For the other 94 percent of the
cases, operation of the makeup pump is sufficient to prevent fuel uncovery.

The calculated fuel uncovery frequency for loss of inventory events is 3.0x10° per year. The
uncovery frequency is low primarily due to the assumption that loss of inventory can drain the
pool only so far. Once that level is reached, additional inventory loss must come from pool
heatup and boiloff. Fuel uncovery occurs if plant fuel handlers fail to initiate inventory makeup
either by use of onsite sources such as the fire pumps or offsite sources such as the local fire
department. In the case of a large leak, isolation of the leak would also be necessary if the
makeup pumps are used. The time available for fuel handler action is considerable, and even in
the case of a large leak, it is estimated that 40 hours will be available. Fuel handlers are alerted
to a loss of inventory condition by control room alarms and indicators, by the visibly dropping
water level in the pool, by the accumulation of water in unexpected locations, and by local
alarms (radiation alarms, building sump high level alarms, etc.).

As with the loss of pool cooling, the frequency of fuel uncovery is calculated to be very low.
Again, a careful and thorough adherence to IDCs #2, #5, #8, and #10 is crucial to establishing
the low frequency. In addition, the assumptions that walkdowns (see SDA #2 above) are
performed on a regular basis (once per shift) and that instrumentation directly measures
temperature and level are important to compensate for potential failures of the instrumentation
monitoring the status of the pool. The assumption that the procedures and/or training give
explicit guidance on the capability of the fuel pool makeup system lowers the expected
probability of fuel handler human errors, and the assumption that fuel handlers will supplement
SFP makeup at appropriate times from alternative higher volume sources lowers the estimated
frequency of failure of the fuel handler to mitigate the loss of coolant inventory. IDCs #6, #7, and
#9 are also credited with lowering the initiating event frequency.

3-14


SEISMIC

Comments #46, 47 and 48: (A) The staff should look at stresses on the transfer tunnel; (B)
Seismic vulnerabilities of the SFP transfer tube should be assessed to properly determine the
tisk of SFP draining; (C) During the July workshop, members of the public raised concerns
about the hazard of the fuel transfer tube interacting with the pool structure during a large
earthquake.

Response: Transfer tubes are generally used in PWR plants where the fuel assembly exits the
containment structure through the tube and enters the pool. These transfer tubes are generally
located inside a concrete structure that is buried under the ground and attached to the pool
structure through a seismic gap and seal arrangement. In most spent fuel pools, the transfer
tube is not connected directly to the area of the pool that contains the spent fuel. The transfer
tube is usually in a separate portion of the pool that has a weir wall separating the area from
the main section that holds the spent fuel. The weir wall is higher than the top of the spent fuel.
As such, even if water was drained through the transfer tube, the fuel would not be uncovered.
Additionally, following the final off-load of the fuel into the spent fuel pool, the transfer tube is
permanently capped at both ends. However, the layouts and arrangements can vary from one
PWR plant to another and the seismic hazard caused by transfer tubes should be examined on
a case-by-case basis. As such, as part of the seismic checklist each licensee must verify the
adequacy of spent fuel pool penetrations whose failure could lead to drainage or siphoning.
(See Appendix 2B)

Comment #49: The staff should address aging effects on the qualification of equipment.

Response: The “equipment qualification” program is required for components that are or have
the potential to be exposed to harsh environments, such as high radiation or high temperature.
Systems around the spent fuel pool are not exposed to harsh environments and therefore do
not needed special consideration. To address the effect of normal aging on spent fuel pool
support systems, the maintenance rule (10 CFR 50.65) requires that the licensee monitor
systems or components associated with the storage, control, and maintenance of the spent fuel
in a safe condition. Additionally, the probability of equipment failure was included in this study
as part of the accident sequences. The staff believes that aging of equipment at
decommissioning plants is adequately addressed through existing programs and characterized
in this study.

Comment #50, 51 and 52: (A) The staff should address aging effects on the spent fuel pool, in
particular, the strengthening or hardening of the concrete and the strength of the liner over
time. (B) The NRC should perform a rigorous engineering analysis of the effects of aging* upon
the spent fuel pool and its associated structures and equipment. Most SFPs were never
designed to be quasi-permanent fuel storage facilities. Because there is, as of yet, no
permanent place to store used fuel, SFPs have had to accept more fuel than they were
originally designed to hold. To allow SFPs to continue to store spent fuel for, as of yet, an
undetermined period of time requires, | suggest a comprehensive look at aging.

(C) A commenter raised concerns about the effect of aging on the spent fuel pool liner plate
and the reinforced concrete pool structure.

* Aging could include degradation, failure, etc. of structures & equipment.

Appendix 6 A6-15


APPENDIX 1.B
TEMPERATURE CRITERIA FOR SPENT FUEL POOL ANALYSIS

1. BACKGROUND

The engineering analyses performed to address spent fuel pool (SFP) performance during
various accidents have, in the past, used a temperature criterion to evaluate the potential for
significant fuel damage. This temperature was intended as an acceptance criterion beyond
which one would expect the onset of significant, global, fuel damage and substantial release of
fission products (e.g., 50-100% of inventory of volatiles) associated with such damage. Further,
the temperature criterion cited (generally about 900°C) has been selected on the basis that it
represented a threshold for self sustained oxidation (Ref. 1) of cladding in air and on that basis it
has been argued that if cooling of the spent fuel could limit fuel temperatures in equilibrium
below this threshold then large releases of fission products need not be considered. Self
sustaining reaction in this sense means the reaction rate and thus heat generation rate is
sufficient, to roughly balance heat losses for given cooling mechanisms, resulting in an
isothermal condition. Once the fuel temperature exceeds this threshold temperature
(alternatively identified as an ignition or autoignition temperature) it was presumed that
subsequent heat up and further increases in reaction rates would be escalating and rapid and
that serious fuel damage would ensue. The temperature escalation associated with oxidation in
this regime would not be balanced by any reasonable cooling afforded by natural circulation of
air. While it was not expected that fission product releases associated with core melt accidents
would immediately emerge at this temperature (based on reactor research in various steam and
hydrogen environments) it was recognized that the time window for subsequent fuel heating
would be relatively small once oxidation escalated. This also did not preclude gap type releases
associated with fuel failures below the threshold temperature but these generally were not
considered to be significant compared to the releases associated with higher fuel temperatures
and significant fuel damage.

In the report, “Draft Final Technical Study of Spent Fuel Pool Accident Risk at Decommissioning
Nuclear Power Plants,” February 2000, the temperature criterion selected, 800 °C, was used in
two ways. First, it was used to determine the decay heat level and corresponding time at which
heat generation and losses for complete and instantaneous draining of the pool would lead to
heating of the fuel (to 800 °C) after 10 hours. This time period would allow for the
implementation of effective emergency response without the full compliment of regulatory
requirements associated with operating reactors. Secondly, the temperature criterion was also
used to evaluate the decay heat level and time (“critical decay time”) at which heat generation
and losses for a fully drained fuel pool would result in an equilibrium temperature of 800 °C.
(typically this critical decay time has been on the order of 5 years). On that basis it was
reasoned that since serious overheating of the fuel had not occurred, the fission product release
associated with core melt need not be considered.

2. AIR OXIDATION AND TEMPERATURE ESCALATION

The NRC has received a number of comments related to the use of this temperature criterion
and has reassessed the appropriateness of such a value for both its intended purposes. At the

Appendix 1B A1B-1


ISSUES OUTSIDE OF THE STUDY

EMERGENCY PREPAREDNESS, SECURITY, INSURANCE, FIRE, AND SAFETY CULTURE

Comment #71: For EP, the integrated decommissioning rule should specify that the licensee is
excused from 10 CFR 50.47 requirements after a period of one-year from final shutdown. The

basis for this recommendation is drawn directly from the technical material presented, and little
can be gained by closer analysis.

Response: The rulemaking will be based on the conclusions drawn from this technical study.
The staff has used all of the information gathered during the technical study to perform its risk
assessment; evaluating the frequencies of events, potential consequences, thermal hydraulic
analysis and the effectiveness of EP.

Comment #72: The decommissioning rule should specify that the licensee is excused from
10 CFR 50.47 off-site EP requirements after the short-lived nuclides important to dose have
undergone substantial decay resulting in off-site dose consequences due to license basis
accidents of less than 1 rem (the EPA protective action guideline).

Response: The staff has considered the decay time of short-lived nuclides and the off-site
dose consequences of the short-lived nuclides. However, to assess the whole risk, the staff
also considered the consequences of longer-lived nuclides, the risks of both design basis
accidents and beyond design basis events, and the effectiveness of EP in efforts to determine
an appropriate point at which requirements for off-site EP could be relaxed.

Comment #73: Section 4.3.2, “Security” of the draft report (February 2000) casts a shadow on
the entire 10 CFR 73.51 rulemaking and needs to clarify the scope of the safety issues. The
last paragraph in Section 4.3.2 should be clear and completely identify the scope and basis of
the ISFSI safety concerns from the radiological sabotage and theft identified in 10 CFR 73.1.
Finally, the last paragraph appears to contradict the May 15, 1998, NRC rulemaking on
Physical Protection for Spent Nuclear Fuel and High-Level Radioactive Waste, Federal
Register Vol. 63, No. 94, Pages 26955 - 26963.

Response: The NRC staff agrees that Section 4.3.2, Security, as written in the February 2000
report, appears to be inconsistent with the changes to Part 73 as described in FRN 26955
dated May 15,1998. The description of risk associated with potential criticality and fuel heat up
is for spent fuel recently discharged from the reactor vessel and not spent fuel stored at an
ISFSI. The discussion is no longer in the report, however, the staff provides additional
information to clarify the apparent inconsistency.

The staff believes that, as written, 10 CFR 73.51 provides proper physical protection for the
storage of spent nuclear fuel licensed under Part 72 at an independent spent fuel storage
installation (ISFSI). The design-basis threat for radiological sabotage of power reactors under
10 CFR 73.1 is not considered appropriate for the types of facilities subject to Section 73.51
and, therefore, a separate protection goal is defined for these facilities. The protection goal
states: "The physical protection system must be designed to protect against loss of control of
the facility that could be sufficient to cause radiation exposure exceeding the dose as described
in 10 CFR 72.106 and referenced by 73.51(b)(3)."

Appendix 6 A6-24


ACKNOWLEDGMENTS

Major portions of the study were done by NRC staff with supporting from subcontractors and
Sandia, Argonne, and Idaho Members. These people are listed below.

Glenn Kelly
Michael Cheok
Gareth Parry
Mark Rubin
Robert Palla
Goutam Bagchi
Robert Rothman
Larry Kopp
Anthony Ulses
Joseph Staudenmeier
Walt Jenson
Christopher Boyd
Jason Schaperow
Charles Tinkler
Sudhamay Basu
Jocelyn Mitchell
Edward Throm
Tanya Eaton
David Diec
Diane Jackson
John Lehning
Paula Magnanelli
United States Nuclear Regulatory Commission

Julie Gregory
Donnie Whitehead
and Randall Gauntt
Sandia National Laboratories

Hee Chung
Argonne National Laboratory

Harold Blackman
Soli Kharicha
Idaho National Engineering and Environment Laboratories

Lawrence Dickson
Atomic Energy of Canada, Limited

Robert Kennedy
RPK Structural Mechanics Consulting, Inc.

Dennis Bley
Buttonwood Consulting


In some cases, emergency preparedness exemptions have also been granted to plants which
were still in the window of vulnerability for zirconium fire. In these cases, the justification was
that enough time had elapsed since shutdown so that the evolution of a zirconium fire accident
would evolve slowly enough to allow mitigative measures and, if necessary, offsite protective
actions to be implemented without preplanning. The staff believes that the technical analysis
discussed in Section 3 and the decision criteria laid out in Section 4.1 provides information on
how such exemption requests could be viewed in the future. In addition, this information bears
on the need for, and the extent of, emergency preparedness requirements in the integrated
rulemaking. In consideration of the study’s conclusion that air cooling may not always be
available for some event sequences, the basis for some previous exemptions may need to be
reconsidered.

The consequence analysis presented in Appendix 4 indicates that the offsite consequences of a
zirconium fire may be comparable to those from operating reactor postulated severe accidents.
Further, the analysis indicates that timely evacuation, implemented through either pre-planned
or ad hoc measures, can significantly reduce the number of early fatalities due to a zirconium
fire. The results in Section 3.7.3 indicate that early fatality and societal risk for an SFP fire may
be comparable to that for an operating reactor, and that the risk is one to two orders of
magnitude lower than the Commission’s safety goal. The results in Appendix 4D show that even
with the most optimistic assumptions about the effectiveness of EP in large seismic events, the
increase in risk associated with relaxations in EP requirements is small and the QHOs continue
to be met. Thus, the risk assessment provides some basis for reductions in EP requirements for
decommissioning plants. With respect to the potential for pool criticality, the staff's assessment
discussed in Section 3 and Appendix 3 indicates that credible scenarios for criticality are highly
unlikely and are further precluded by the assumption of Boraflex monitoring programs.
Additionally, even if a criticality event did occur, it would not have offsite consequences.
Therefore, the conclusions regarding possible reductions in EP program requirements are not
affected.

In Section 4.1, the safety principles of RG 1.174 are applied to assess whether changes to
emergency preparedness requirements are appropriate. Notwithstanding the low risk
associated with SFP accidents, the safety principles in RG 1.174 dictate that defense-in-depth
be considered. As discussed previously, emergency preparedness provides defense-in-depth.
However, because of the considerable time available to initiate protective actions, in most SFP
accidents (and the low frequency of events in which sufficient time is not available to implement
protective actions on an ad hoc basis), the level of formal emergency plans needed for rapid
initiation and implementation of offsite protective actions can be evaluated. The principle
emergency planning measures needed for SFP events is the means for identifying the event
and notifying of State and local emergency response officials.

4.2.2 Security

Currently licensees that have permanently shut down reactor operations and have off-loaded
the spent fuel into the SFP are still required to meet all the security requirements for operating
reactors in 10 CFR 73.55 (Ref. 9). This level of security requires a site with a permanently
shutdown reactor to provide security protection at the same level as for an operating reactor
site. The industry has asked the NRC to consider whether the risk of radiological release from

4-14


6.0 REFERENCES

References for Executive Summary

1.

U.S. Code of Federal Regulations, “Domestic Licensing of Production and Utilization
Facilities,” Part 50, Title 10, “Energy.”

U.S. Nuclear Regulatory Commission, “An Approach for Using Probabilistic Risk
Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,”
Regulatory Guide 1.174, July 1998.

Dana A. Powers, Chairman of the Advisory Committee on Reactor Safeguards, U.S.
Nuclear Regulatory Commission, letter to Dr. William D. Travers, U.S. Nuclear Regulatory
Commission, “Spent Fuel Fires Associated With Decommissioning,” November 12, 1999.

U.S. Code of Federal Regulations, “Standards for Protection Against Radiation,” Part 20,
Title 10, “Energy.”

References for Section 1.0

1.

U.S. Code of Federal Regulations, “Domestic Licensing of Production and Utilization
Facilities,” Part 50, Title 10, “Energy.”

References for Section 2.0

1. U.S. Nuclear Regulatory Commission, "Use of Probabilistic Risk Assessment Methods in
Nuclear Regulatory Activities; Final Policy Statement," Federal Register, Vol. 60, No. 158,
August 16, 1995.

2. U.S. Nuclear Regulatory Commission, “An Approach for Using Probabilistic Risk
Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,”
Regulatory Guide 1.174, July 1998.

3. Dana A. Powers, Chairman of the Advisory Committee on Reactor Safeguards, U.S.
Nuclear Regulatory Commission, letter to Dr. William D. Travers, U.S. Nuclear Regulatory
Commission, “Spent Fuel Fires Associated With Decommissioning,” November 12, 1999.

4. U.S. Code of Federal Regulations, “Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants,” Section 65, Part 50, Title 10, “Energy.”

References for Section 3.0

1. U.S. Nuclear Regulatory Commission, “Accident Source Terms for Light Water Nuclear
Reactor Power Plants,” NUREG-1465, February 1995.

2. U.S. Nuclear Regulatory Commission, “Regulatory Analysis for the Resolution of Generic

Issue 82, ‘Beyond design Basis Accidents in Spent Fuel Pools,”” NUREG-1353, April 1989.

6-1


frequency of load drops with the conditional probability of pool failure given a load drop, the staff
was able to estimate the frequency of heavy load drops causing a zirconium fire at
decommissioning facilities.

3. | FREQUENCY OF HEAVY LOAD DROP

The database used in this evaluation (primarily the 1990s Navy data) considered a range of
values for the number of occasions where faults occurred, the frequency of heavy load drops
and the availability of backup systems. The reason that there is a range of values is that while
the number of equipment failures and load drops were reported, the denominator of the
estimate, the actual total number of heavy load lifts, was only available based on engineering
judgement. High and low estimates of the ranges were made, and it was assumed that the data
had a log normal distribution with the high and low number of the range representing the 5" and
95" percentile of the distribution. From this the mean of the distribution was calculated. Data
provided by NEI on actual lifts and setdowns of SFP casks at commercial U.S. nuclear power
plants (light water and gas-cooled reactors) gave a similar estimated range for the incidents at
the 95 percent confidence level.

Load drops were broken down into two categories: failure of lifting equipment and failure to
secure the load.

Crane failures (failure of lifting equipment) were evaluated using the fault tree shown in

Figure A2c-1, which comes from NUREG-0612. At the time that heavy loads were evaluated in
NUREG-0612, low-density storage racks were in use and after 30 to 70 days (a period of about
0.1 to 0.2 per year), no radionuclide releases were expected if the pool were drained. It was
assumed in NUREG-0612 that after this period, the fuel gap noble gas inventory had decayed
and no zirconium fire would have occurred. Today, most decommissioning facilities use high-
density storage racks. This analysis evaluates results at one year after reactor shutdown. Our
engineering evaluations indicate that for today’s fuel configurations, burnup, and enrichment, a
zirconium cladding fire may occur if the poo! were drained during a period as long as 5 years.

A literature search performed by the staff searching for data on failure to secure loads identified
a study (WIPP report) that included a human error evaluation for improper rigging. This study
was used by the staff to re-evaluate the contribution of rigging errors to the overall heavy load
(cask) drop rate and to address both the common mode effect estimate and the 1990s Navy
data. Failure to secure a load was evaluated in the WIPP report for the Trudock crane. The
WIPP report determined that the most probable human error was associated with attaching the
lifting legs to the lifting fixture. In the WIPP report, the failure to secure the load (based on a
2-out-of-3 lifting device) was estimated based on redundancy, procedures, and a checker. The
report assumed that the load could be lowered without damage if no more than one of the three
connections were not properly made. Using NUREG/CR-1278 (Ref. 4) information, the mean
failure rate because of improper rigging was estimated in the WIPP report to be 8.7x107 per lift.
Our requantification of the NUREG-0612 fault tree using the WIPP improper rigging failure rate
is summarized in Table A2c-2. The WIPP evaluation for the human error probabilities is
summarized in Table A2c-3.

These estimates provided a rate for failures per lift. Based on input from the nuclear industry at

the July 1999 SFP workshop, we assumed in our analysis that there will be a maximum of
100 cask lifts per year at a decommissioning plant.

Appendix 2C. A2C-2


Comment #56: The NRC should specify why it is not cost effective to perform a plant-specific
seismic evaluation for each spent fuel pool and what impact this has on safety. Because there
are so many differently designed spent fuel pools, it is difficult to perceive how a generic
approach could be acceptable without assembling a list of similar and/or identical designs and
performing a seismic evaluation of the various groups which are assembled. Specific seismic
evaluations for each plant or groups of similar/identical plants should be considered.

Response: A significant body of work exists characterizing the strength and capacity of shear
walls based on tests and analyses. The use of a generic parameter, with the underpinning of
data, solely for the purpose of screening is very appropriate and reliable. Using the seismic
checklist, a structure is not acceptable unless all the conditions in the checklist are met. At
sites where the seismic checklist is not met, a plant specific evaluation can be conducted. The
use of a screening parameter is a reliable way to determine the need for further evaluation.

Comment #57: Did the NUREGs that you looked at take into account new information coming
out of the Kobe and Northridge events? Particularly as we are learning more about risks
associated with those two particular seismological events that were never even considered
when plants were sited; particularly, though | can’t frame it in the seismological language, from
a lay understanding, it’s clear that new information was gained out of Kobe and Northridge
events suggesting that you can have seismological effects of greater consequence farther
afield than at the epicenter of the event.

Response: The staff believes that the NUREG reports mentioned by the commenter were
NUREG 1488, “Revised Livermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites
East of the Rocky Mountains” and NUREG/CR-5250, “Seismic Hazard Characterization of 69
Nuclear Plant Sites East of The Rocky Mountains,” which were written in the middle and late
1980s and used probabilistic seismic hazard analyses performed for the NRC by Lawrence
Livermore National Laboratory (LLNL) for nuclear power plants in the central and eastern U.S.
Since then, LLNL has performed additional probabilistic hazard studies (1993) for central and
eastern U.S. nuclear power plants for the NRC. The results of the more recent study indicated
lower seismic hazards for the plants than the earlier study estimated. EPRI has also performed
probabilistic hazard studies. The LLNL hazard curves generally predict higher frequency
estimates than those generated by EPRI. This is a result of different expert judgements.
However, both are valid methodologies.

The design basis for each nuclear power plant took into account the effects of earthquake
ground motion. The seismic design basis, called the safe shutdown earthquake (SSE), defines
the maximum ground motion for which certain structures, systems, and components necessary
for safe shutdown were designed to remain functional. The licensees were required to obtain
the geologic and seismic information necessary to determine site suitability and provide
reasonable assurance that a nuclear power plant could be constructed and operated at a site
without undue risk to the health and safety of the public.

The information collected in the investigations was used to determine the earthquake ground
motion at the site, assuming that the epicenters of the earthquakes are situated at the point on
the tectonic structures or in the tectonic provinces nearest to the site. The earthquake which
could cause the maximum vibratory ground motion at the site was designated the SSE. This
ground motion was used in the design and analysis of the plant.

The determination of the SSEs followed the criteria and procedures required by NRC
regulations and applied a multiple hypothesis approach. In this approach, several different

Appendix 6 A6-18


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Appendix 4D


The appropriateness of the PPG was questioned by the ACRS in view of potential effects of the
fission product ruthenium, the release of fuel fines, and the effects of revised plume parameters.
The staff added sensitivity studies to its analyses to examine these issues. The consequences
of a significant release of ruthenium and fuel fines were found to be notable, but not so important
as to render inappropriate the staff's proposed PPG of 1x10° per year. The plume parameter
sensitivities were found to be of lesser significance.

In its thermal-hydraulic analysis, documented in Appendix 1A, the staff concluded that it was not
feasible, without numerous constraints, to establish a generic decay heat level (and therefore a
decay time) beyond which a zirconium fire is physically impossible. Heat removal is very
sensitive to these additional constraints, which involve factors such as fuel assembly geometry
and SFP rack configuration. However, fuel assembly geometry and rack configuration are plant
specific, and both are subject to unpredictable changes after an earthquake or cask drop that
drains the pool. Therefore, since a non-negligible decay heat source lasts many years and
since configurations ensuring sufficient air flow for cooling cannot be assured, the possibility of
reaching the zirconium ignition temperature cannot be precluded on a generic basis.

The staff found that the event sequences important to risk at decommissioning plants are limited
to large earthquakes and cask drop events. For emergency planning (EP) assessments this is
an important difference relative to operating plants where typically a large number of different
sequences make significant contributions to risk. Relaxation of offsite EP a few months after
shutdown resulted in only a “small change” in risk, consistent with the guidance of RG 1.174.
Figures ES-1 and ES-2 illustrate this finding. The change in risk due to relaxation of offsite EP is
small because the overall risk is low, and because even under current EP requirements, EP was
judged to have marginal impact on evacuation effectiveness in the severe earthquakes that
dominate SFP risk. All other sequences including cask drops (for which emergency planning is
expected to be more effective) are too low in likelihood to have a significant impact on risk. For
comparison, at operating reactors additional risk-significant accidents for which EP is expected
to provide dose savings are on the order of 1x10° per year, while for decommissioning facilities,
the largest contributor for which EP would provide dose savings is about two orders of
magnitude lower (cask drop sequence at 2x10” per year).' Other policy considerations beyond
the scope of this technical study will need to be considered for EP requirement revisions and
previous exemptions because a criteria of sufficient cooling to preclude a fire cannot be satisfied
on a generic basis.

Insurance does not lend itself to a “small change in risk” analysis because insurance affects
neither the probability nor the consequences of an event. As seen in figure ES-2, as long as a
zirconium fire is possible, the long-term consequences of an SFP fire may be significant. These
long-term consequences (and risk) decrease very slowly because cesium-137 has a half life of
approximately 30 years. The thermal-hydraulic analysis indicates that when air flow has been
restricted, such as might occur after a cask drop or major earthquake, the possibility of a fire
lasts many years and a criterion of “sufficient cooling to preclude a fire” can not be defined on a

‘Consistent with PRA limitations and practice, contributions to risk from safeguards
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Appendix 4D


temperature, airborne radiation), or provide indirect evidence (sump pump alarms,
secondary side cooling system trouble alarms)

The first and last of these factors may be reflected in the reliability assumed for the alarm and in
the structure of the logic model (fault tree) for the event tree function control room alarms (CRA),
respectively. The other factors may be taken into account in assessing the reliability of the
operator response.

For active monitoring, examples of the factors used in assessing the effectiveness of the
monitoring include:

. scheduled walk-downs required within areas of concern, with specific items to check
(particularly to look for indications not annunciated in, or monitored from, the control room,
for example, indications of leakage, operation of sump pumps if not monitored, steaming
over the pool, humidity level)

° plant operating procedures that require the active measurement of parameters (e.g.,
temperature, level) rather than simply observing the condition of the pool

. requirement to log, check, and trend results of monitoring
. alert levels specified and noted on measurement devices

These factors can all be regarded as performance shaping factors (PSFs) that affect the
reliability of the operators.

An important factor that should mitigate against not noticing a deteriorating condition is the time
scale of development, which allows the opportunity for several shifts to notice the problem. The
requirement for a formal shift turnover meeting should be considered.

3.2.3.2 Situation Assessment and Response Planning

The principal operator aids for situation assessment and response planning are procedures and
training in their use.

The types of procedures that might be available are:

. annunciator/alarm response procedure that is explicit in pointing towards potential
problems

. detailed procedures for use of alternate systems indicating primary and back-up sources,
recovery of power, etc.

The response procedures may have features that enhance the likelihood of success, for
example:

. inclusion of guidance for early action to establish contingency plans (e.g., alerting offsite

Appendix 2A A2A-7


concurrently with the earthquake should also be considered if the events contribute to the worst
case scenario.

Response: Risk assessments are performed as realistic as possible, not worst case. The staff
identified the following nine initiating event categories to investigate as part of the quantitative
risk assessment on SFP risk:

Loss of Off-site Power from plant centered and grid related events
Loss of Off-site Power from events initiated by severe weather
Internal Fire

Loss of Pool Cooling

Loss of Coolant inventory

Seismic Event

Cask Drop

Aircraft Impact

Tornado Missile

The initiating events indicated above are independent. However, the event sequences that
emanate from each event are carefully modeled in the event tree and could include some of the
same circumstances. This means that a seismic event tree would include the consideration of
off-site and on-site power loss. In a PRA assessment no risk insight can be gained by
considering worst case combination of truly random and independent events such as a seismic
event and a tornado missile. However, the frequency of a combined seismic and tornado
missile is much less than 1x10°. Also, with respect to other structures, such as crane girders
and super-structures, they are covered in the seismic check list for the spent fuel pool
structure.

Comment #55: The NEI seismic checklist requires a seismic engineer to review drawings in
addition to conducting a walkdown of the SFP. It has been my experience that many electrical
drawings of NPP’s do not reflect the existing plant electrical installation. How is the seismic
engineer going to verify drawings to the existing SFP building and pool if much of the pool is
inaccessible? For instance, how does he verify concrete degradation under the steel liner?
The NRC should require that specific areas be inspected and that these areas be accessible. If
these areas are not accessible, then the checklist is not complete and susceptibility to seismic
activity remains a concern.

Response: Plant walkdowns should provide an opportunity to verify and correct plant drawings
to the as-built conditions, and design calculation would verify the compliance to Codes and
Standards. The staff considers the review of construction drawings to be very important.
Minimum reinforcing areas are dictated by codes. Thick walls and slabs forming spent fuel pool
structure are in many cases governed by minimum reinforcing requirements. Should there be
any additional shear or flexural steel requirements, engineering calculations would indicate
where they are needed and how much is needed. Therefore, a review of drawings and design
calculations would present a more complete picture. With respect to inaccessible areas,
cracks, spalling of concrete and stains, and efflorescence are of a degradation process. In
order to determine the root cause of the external signs, it is necessary to use more invasive
procedures, such as chipping and breaking concrete, etc. This is not unique to spent fuel pool
structures, and there are several examples of this type of inspection in the operating
experience of several plants.

Appendix 6 A6-17


The appropriateness of the PPG was questioned by the ACRS in view of potential effects of the
fission product ruthenium, the release of fuel fines, and the effects of revised plume parameters.
The staff added sensitivity studies to its analyses to examine these issues. The consequences
of a significant release of ruthenium and fuel fines were found to be notable, but not so important
as to render inappropriate the staffs proposed PPG of 1x10° per year. The plume parameter
sensitivities were found to be of lesser significance.

In its thermal-hydraulic analysis the staff concluded that it was not feasible, without numerous
constraints, to establish a generic decay heat level (and therefore a decay time) beyond which a
zirconium fire is physically impossible. Heat removal is very sensitive to these additional
constraints, which involve factors such as fuel assembly geometry and SFP rack configuration.
However, fuel assembly geometry and rack configuration are plant specific, and both are subject
to unpredictable changes after an earthquake or cask drop that drains the pool. Therefore,
since a non-negligible decay heat source lasts many years and since configurations ensuring
sufficient air flow for cooling cannot be assured, the possibility of reaching the zirconium ignition
temperature cannot be precluded on a generic basis.

The staff found that the event sequences important to risk at decommissioning plants are limited
to large earthquakes and cask drop events. For emergency planning (EP) assessments this is
an important difference relative to operating plants where typically a large number of different
sequences make significant contributions to risk. Relaxation of offsite EP a few months after
shutdown resulted in only a “small change” in risk, consistent with the guidance of RG 1.174.
The change in risk due to relaxation of offsite EP is small because the overall risk is low, and
because even under current EP requirements, EP was judged to have marginal impact on
evacuation effectiveness in the severe earthquakes that dominate SFP risk. All other
sequences including cask drops (for which emergency planning is expected to be more
effective) are too low in likelihood to have a significant impact on risk. For comparison, at
operating reactors additional risk-significant accidents for which EP is expected to provide dose
savings are on the order of 1x10° per year, while for decommissioning facilities, the largest
contributor for which EP would provide dose savings is about two orders of magnitude lower
(cask drop sequence at 2x10” per year). Other policy considerations beyond the scope of this
technical study will need to be considered for EP requirement revisions and previous
exemptions because a criteria of sufficient cooling to preclude a fire cannot be satisfied on a
generic basis.

Insurance does not lend itself to a “small change in risk” analysis because insurance affects
neither the probability nor the consequences of an event. The study found that as long as a
zirconium fire is possible, the long-term consequences of an SFP fire may be significant. These
long-term consequences (and risk) decrease very slowly because cesium-137 has a half life of
approximately 30 years. The thermal-hydraulic analysis indicates that when air flow has been
restricted, such as might occur after a cask drop or major earthquake, the possibility of a fire
lasts many years and a criterion of “sufficient cooling to preclude a fire” can not be defined on a
generic basis. Other policy considerations beyond the scope of this technical study will
therefore need to be considered for insurance requirements.

The study also discusses implications for security provisions at decommissioning plants. For
security, risk insights can be used to determine what targets are important to protect against

5-2


LOSS EXCEEDS NORMAL,

CAPACITY AND

ISOLATION OF BREAK IS
NECESSARY

PROBABLITY THATA
LOLEVENT DOES NOT
EXCCED MAKEUP

LoGuK

OPERATORFALS
RESPOND
CRALARM

Lov

OPERATOR FALS TO
RESPOND ALARM

HEP. DIAG ALARM

A-15

LOCAL,
FAULTS CASING
LOSS OF INDICATION

SPCLVL-LOP



pumps, a heat exchanger, an ultimate heat sink, a makeup tank, filtration system and isolation
valves. Suction is taken via one of the two pumps on the primary side from the SFP and is
passed through the heat exchanger and returned back to the pool. One of the two pumps on
the secondary side rejects the heat to the ultimate heat sink. A small amount of water is
diverted to the filtration process and is returned to the discharge line. A regular makeup system
supplements the small losses because of evaporation. In the case of prolonged loss of SFPC
system or loss of inventory events, the inventory in the pool can be made up using the firewater
system. There are two firewater pumps, one motor-driven (electric) and the other diesel-driven,
which provide firewater throughout the plant. A firewater hose station is provided in the SFP
area. The firewater pumps are assumed to be located in a separate structure.

3.0 METHODOLOGY

3.1 Logic Model

This section summarizes the SFP PRA model developed in this study. The description of the
modeling approach and key assumptions is intended to provide a basis for interpreting the
results in Sections 4 and 5. The event trees and fault trees presented in this study are meant to
be generic enough to apply to many different configurations. The fault trees are documented in
Attachment A to this appendix. An example of the HRA worksheet used for this analysis is
presented in Attachment B.

The endstate for this analysis is defined as loss of coolant inventory to the point of fuel uncovery
from either leakage or boil-off. Dose calculations (Ref. 5) show that when there is less than

3 feet of water above the top of the fuel, an environment that is rapidly lethal to anyone at the
edge of the pool can result. Therefore, 3 feet has been adopted as an effective limit for recovery
purposes. In other words, the endstate for this analysis is effectively defined as loss of coolant
inventory to a point 3 feet above the top of the fuel. One of the NEI commitments is that there
should be a provision for remote alignment of the makeup source to the pool, which would make
this assumption conservative. However, the impact of this conservatism on the conclusions of
this analysis is minor.

The event tree and fault tree models were developed and quantified using Version 6 of the

SAPHIRE software package (Ref. 6), using a fault tree linking approach. Event trees were
developed for each of the initiators identified in Section 1.

3.2 HRA Methodology

3.2.1 Introduction

One of the key issues in performing a probabilistic risk assessment (PRA) for the SFP during the
decommissioning phase of a nuclear power plant's life cycle is how much credit can be given to
the operating staff to respond to an incident that impacts the SFP that would, if not attended to,
lead to a loss of cooling of the spent fuel and eventually to a zirconium fire.

The objective of the HRA analysis in this PRA is to assess whether the design features and
operational practices assumed can be argued to suggest that the non-response probabilities

Appendix 2A A2A-3


and the fuel allow for simple estimation of the time it would take to reach temperatures at which
significant zirconium oxidation would begin or significant fission product releases would occur.

Figure 1A-2 shows a comparison of the air-cooled calculation to an adiabatic heatup calculation
for a PWR at a burnup of 60 GWD/MTU. The timing of the event is important because of the
infinite number of configurations that are possible after a dynamic event. The results show that
the air-cooled heatup times are shorter than the adiabatic heatup times for times up to 2 years
after shutdown. This is because the air cooling heatup rate is close to the adiabatic heatup rate
and the oxidation heat source becomes a significant contributor to the total power at
temperatures of approximately 600 °C. An adiabatic heatup calculation that included the
oxidation heat source would have heatup times shorter than the air-cooled heatup times. The
results show a heatup time to fission product release of 4 hours at 1 year after shutdown for a
PWR with 60 GWD/MTU fuel burnup even with unobstructed airflow. At 5 years after shutdown
the release of fission products may occur approximately 24 hours after the accident even with
obstructed airflow. The air-cooled calculations used the parabolic oxidation rate equation
recommended in Appendix 1B. This oxidation model leads to the fastest heatup times of the
oxidation models that were examined.

The staff attempted to calculate a generic critical decay time necessary to ensure that air cooling
was adequate to prevent the clad temperature from reaching the temperature of self-sustaining
zirconium oxidation. The staff determined that it was not feasible absent setting stringent
requirements or restrictions on plant fuel rack configurations, fuel burnup, and building
ventilation to calculate a generic critical decay time. The staff examined the impact of bumup,
oxidation models, building ventilation volumetric flow rate, and downcomer flow on the critical
decay time. Burnup and variation in oxidation models are of secondary importance relative to
air flow. As noted above air flow is dependent upon factors which can vary widely from pool to
pool and even within an SFP. Therefore, the staff used an unobstructed flow model for one
bound, and an adiabatic model as a second bound for its analyses. As seen in Appendix 1.B,
the maximum clad temperature that is used for the definition of the critical decay time is
dependent on the time after shutdown. The maximum clad temperature allowed is 600 °C for
times less than 5 years after shutdown and 800 °C for times greater than 5 years after
shutdown. Ultimately, the time differences resulting from these temperatures are not significant
because of the already short time available in the early years where the source term is changing
significantly because of ruthenium decay.

Appendix 1A A1A-5


PROS AARE
FALSSPC
‘SYSTEM
HEPRESARE
L .
FAILURE TO MAKELP-
INVENTORY USING
DGRIRE PUMP
Fi
n
i t 1
OPERATOR FALS
DIESEL PUP FAILS FAILS TO DIAGNOSE
‘TO ESTABUSH
ALTERNATE COCLING. TOFUNCTION LOSS OF SFP COQLING
~ EARLY
HEP-AWSTART HEP-RECG PASTART
FAILS TOREPAR
Bese. WPLNE pRoTecTON uve
OePENCE! FALSSTART,
any START AND RUN
HEP-FWREP-NODEP FP.OGPUNP-FTF

A-7



sero mato vane eer
SORE ‘REET | oe SE.
' t
ZEROS EM TO OFERATORFALS TO Sas ORR, ESE AD BECTRC
Steen Sess Oe See SRE
wode Bd, abedn Q.
t :
FALURETO MAKEUP
INVENTORY USING
DGFIRE PUP
aL
J
f t 1
pes csuCatierag GIESEL FIRE PROTECTION FAILS TO RECOGNIZE
COOLING USING FW PUMP FAILS TO BROM ‘NEED TOSTART FW
HEP-AWSTART GLPRIAZLPLA HEP-RECGAASTART
[ 1 1
DIESEL FIRE PROTECTION FALS TOREPAR
PUMP FAILS TOSTART RWOESE: MP
AND RUN (85 HRS)
FP.DGPUNP-FTF HEP-FWREP-NODEP

A-10



* These observations are valid regardless of whether seismic event frequencies are based
on the LLNL or the EPRI seismic hazard estimates.

Figures 3.7-7 and 3.7-8 show the risk measures relevant to the Commission's safety goal policy
statement, specifically, the individual risk of early fatality (to an individual within 1 mile of the site)
and the individual risk of latent cancer fatality (to an individual within 10 miles of the site). The
upper curves are based on the LLNL seismic hazard curves and the high ruthenium source
term, and the lower curves are based on the EPRI hazard curves and the low ruthenium source
term. Accordingly, these results may be viewed as a representative range of risk results for
spent fuel pools uncovery given the conservative assumption that all SFP accidents result in a
fire.

3-46


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as

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WIM Woyog Yee

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OSLL-DSYNN
woy synsey
Jojoeay BueredQ

daiing

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ySIy (Wias-UOSI9d) [2}a190g [OO Jen4 jueds

OL

ool

(aeak Jad) Say OS UlUPAA Wes-UOS19g

3-42


Figure 3.1 Assumed Spent Fuel Pool Cooling System

Makeup {
Tank «| *Fuel Pool
Makeup Pump

Hose -
i | —é
ay ea!
nee a arene |
| i | MP2 |
Spent Fuel Pool =| bg D4 !
ae (SFP) = ~MP-1
fo | Fuel Pool '
i ————— Cooling Pumps _———, |
Fuel Pool .
He h.
leat Exct
Diesel-Driven| a
| OSG i |
| Fleotic | Reser OO,
Se — Li
ican | i 7
Fire Pum ~ Ultimate
ia Heat Sink i
(Ait Coolers) |

ccwosra

Based upon information obtained during the site visits and discussions with decommissioning
plant personnel during those visits, the staff also made the following assumptions that are
believed to be representative of a typical decommissioning facility:

+ The SFP cooling design, including instrumentation, is at least as capable as that assumed
in the risk assessment. Licensees have at least one motor-driven and one diesel-driven fire
pump capable of delivering inventory to the SFP (SDA #1, Table 4.2-2).

+ The makeup capacity (with respect to volumetric flow) is assumed to be as follows:

Makeup pump: 20 - 30 gpm

Firewater pump: 100 — 200 gpm
Fire engine: 100 — 250 gpm (100 gpm, for hose: 1%-in., 250 gpm for 21/2-in. hose)

3-4


From
Sheet 1

Figure A2c-1 (sheet 2 of 2) - Heavy Load Drop Fault Trees

LOI-S

LOSS-OF-INVENTORY DUE TO
FAILURE OF HEAVY LOAD
(CRANE AND RIGGING) SYSTEM

RF CFCR [ i
vervmich ese] | Tt elues (crane and gana) Frain ofpain | | Frachon ofpat
i
May occur P Near/over pool oad drop
CF I é; cR
Failure of crane Failure of rigging
Toad Sop Leading to
CRANE D1 RIGGING o2
FAILURE Lift per year Lift per year
OF Leading to drop FAILURE Leading to drop
CRANE (crane related) feromn k RIGGING {rigging related)

A2C-10


z-L'€ eunBi{

UMOPINYS [eUly Aa SYUOW

0zL 80L 96 v8 eZ 09 8b 9¢ ve et 0
fl I | 1 ' 1 a a A —+ go+at
L uonenseag Ayleg
(9+39'p) Aung % ‘uua] eounog wintusyjny MOT
do} UEP|ooy PayeHiuy 7 S65
Ajjeusezuy ys10\\ ——¥- Y- ¥.
+ uojenseng e127 1
(242 }) Aung y,
-L We) BouNoS WI MO"
JO} JUBPIODY payeniul a | BOuNOS WiNUEYINY MO?
Ajpeoiwsiag ysioy, - |
( ) ~ ~ Ll * ; 2O+31
2432 b uojenoeng Ayeq a
woHog YOESd ? >: ‘wJ84 eounog wnueyyNY YB ba |
Jo} juapiooy payeniuy °.
Apeuseyul SION, >
[ uoyenoenq oye] +, ‘e!
(24361) ‘wal 8ounog winjueuyny UB
woyog yoeeq | +.
Jo} uaplooy payeniul
AWEDIWSIOS JSIONA, +
OSLt-O3YNN +r
Woy s}INsey +
Joyoeay Buyeiedo T |
80+31

SU9 | 89.1NOS JOO [any juads 10} SadUsNbasuUOD (WeJ-UOSJed) |e}9190S

(quaas sad) say OG UIY}IAA Wds-UOS19g

3-33


Figure 3

Appendix 3

LEGEND
voID

WM MATERIAL 1

[S@I MATERIAL 2

TEBE MATERIAL 3

. MATERIAL 4

WM MATERIAL 6

Sample Geometry Assuming 4 Assembly Spacing Between Most Reactive

Assembly

A3-16



EP case and minimizes its effectiveness in the relaxed preplanning case, thereby tending to
maximize the risk increase associated with EP relaxations.

The estimated risk increases associated with the EP relaxation are summarized in Table 4.
The results indicate that relaxation of the requirements for radiological preplanning would result
in an increase of about 1.5x10° early fatalities per year and 2 person-rem per year, which is
about a factor of 15 and five below the allowable increase inferred from the RG 1.174 LERF
criteria. The other risk measures are also substantially lower than the allowables from RG
1.174. Since the SFP fire frequency assumed in these comparisons (2.4x10°* per year) is
about a factor of four lower than the PPG of 1x10° per year, a plant operating nominally at the
PPG would have a smaller margin to the allowable risk limits for the reference plant but would
still be at or below the limits under the above assumptions.

The results of the sensitivity studies indicate that even under unrealistically optimistic
assumptions regarding the value of EP in seismic events, the change in risk associated with
relaxation of the requirements for radiological preplanning is still relatively small. The increases
in early fatalities and individual early fatality risk remain below the maximum allowable for each
tisk measure. Population dose and individual latent cancer fatality risk are about a factor of two
higher than the allowable value inferred from RG 1.174. However, this increase in individual
latent cancer risk represents only about 9 percent of the QHO, and considerable margin to the
QHO would still remain.

It must be kept in mind that the evacuation effectiveness assumed for "Full EP” in the sensitivity
case is unrealistic for high g earthquakes, and that the risk increase associated with the EP
relaxations would be closer to the baseline value. Also, the risk reduction estimates are based
on the LLNL seismic hazard frequencies and the high ruthenium source term, and would be
substantially lower if either the EPRI seismic hazard frequencies or the low ruthenium source
term were used. Finally, the above comparisons are based on the risk levels 1 year after
shutdown.

The impact of the above factors on the maximum risk increase for the EP relaxations is shown
in Figures 1 and 2 for early fatalities and population dose (person-rem). Use of either the EPRI
seismic hazard frequencies or the low ruthenium source term would reduce each of the risk
measures by about a factor of 10, to values which are well below the RG 1.174 allowables.
The risk impact will decrease in later years because of reduced consequences as fission
products decay further.

Appendix 4D A4D-3


i.e., a frequency of 2x10° per year to reflect the use of the LLNL hazard estimates and a
frequency of 2x10” per year to reflect use of the EPRI hazard estimates. Use of the mean value
facilitates comparisons with the Commission’s quantitative safety goals and quantitative health
objectives (QHOs). About 70 percent of the sites are bounded using the mean value.

Table 3.7-4 Mean Frequency of Rapid Draindown Due to Seismic Events

Source of Hazard Estimate Frequency (per year)
LENL 2x10
EPRI 2x10?

Likely SFP failure modes and locations are discussed in Attachment 2 to Appendix 2B. The
conclusion is that drainage of the pool would be fairly rapid and a small amount of water is likely
to remain in the pool, with post-seismic-failure, water depths ranging from about zero to about

4 feet depending upon the critical failure mode. For purposes of consequence assessment, all
seismically initiated sequences were assumed to result in a rapid draindown followed by an SFP
fire, regardless of the SFP failure mode and location, which are plant-specific.

The SFP risk estimates are strongly dependent on the assumptions about the effectiveness of

emergency evacuation in seismic events, since these events dominate the SFP fire frequency.

In NUREG-1150, evacuation in seismic events was treated in either of two ways, depending on
the peak ground acceleration (PGA) of the earthquake:

+ For low PGA earthquakes, the population was assumed to evacuate; however, the
evacuation was assumed to start later and proceed more slowly than evacuation for
internally initiated events.

* For high PGA earthquakes, it was reasoned that there would be no effective evacuation
and that many structures would be uninhabitable.

Since the seismic contribution to SFP fire frequency is driven by events with ground motion
several times larger than the SSE, the reasoning that there would be no effective evacuation
was adopted in developing the seismic contribution to the risk. This is consistent with the expert
opinion provided in Attachment 2 to Appendix 2B about the expected level of collateral damage
within the emergency planning zone in a seismic event large enough to cause the SFP failure.
Specifically, for ground motion levels that correspond to SFP failure in the Central and Eastern
United States, it is expected that electrical power would be lost and more than half of the bridges
and buildings (including those housing communication systems and emergency response
equipment) would be unsafe even for temporary use within at least 10 miles of the plant. This
approach is also consistent with previous Commission rulings on San Onofre and Diablo
Canyon in which the Commission found that for those risk-dominant earthquakes that cause
very severe damage to both the plant and the offsite area, emergency response would have
marginal benefit because of offsite damage.

3-37


Boraflex degradation is another problem that is somewhat unique to BWR spent fuel storage
racks. This is true because of the fact that BWR storage pools do not contain soluble boron
that provides the negative reactivity in PWR pools to offset the positive effect of Boraflex
degradation. Therefore, some compensatory measures need to be in place to provide
adequate assurance that Boraflex degradation will not contribute to a criticality event. In
operating reactor spent fuel pools that use Boraflex, licensees use some sort of surveillance
program to ensure that the 5 percent subcritical margin is maintained. These programs should
be continued during and following decommissioning. No criticality calculations were performed
for this study to assess Boraflex degradation because it is conservatively assumed that the loss
of a substantial amount of Boraflex will most likely lead to a criticality accident.

These analyses are not all inclusive, but we believe that the physical controls identified are
generically applicable. We examined all of the available GE designed BWR assemblies for
which information was available and identified the assembly used in the study to have the
largest K,, in the standard cold core geometry (in other words, in the core with no control rods
inserted at ambient temperature) at the time of peak reactivity. This assembly was a GE12
design (10x10 lattice) enriched to an average value of 4.12 w/o Uz35._ Only the dominant part of
the lattice was analyzed and it was assumed to span the entire length of the assembly. This
conservatism plus the fact that the assembly itself is highly enriched and designed for high
burnup operation has led the staff to conclude that these analyses are generically applicable to
BWR spent fuel storage pools.

Conclusions

One scenario that has been identified which could lead to a criticality event is a heavy load drop
or some other event that compresses a low density rack filled with spent fuel at its peak
expected reactivity. This event is somewhat unique to decommissioned reactors because
there are more low burnup (high reactivity) assemblies stored in the spent fuel pool that were
removed from the core following its last cycle of operation, than in a SFP at an operating plant.

To address the consequences of the compression of a low density rack, there are two
strategies that could be used, either individually or in combination. First, the most reactive
assemblies (most likely the fuel from the final cycle of operation) could be scattered throughout
the pool, or placed in high density storage if available. Second, all storage pools, regardless of
reactor type, could be borated.

References

1. “ENDF/B-V Nuclear Data Guidebook," EPRI-NP 2510, July 1982.
2 "SCALE: A Modular Code System for Performing Standardized Computer Analyses for
Licensing Evaluations," NUREG/CR-0200. Oak Ridge National Laboratory, 1995.

3. Tony Ulses, "Evaluation of NEWT for Lattice Physics Applications," Letter Report, May
1999. :

4 M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-Group
ENDF/B-V Cross Section Library for use in Criticality Safety Analysis,” NUREG/CR-6102,
Oak Ridge National Laboratory, 1994.

5  O.W. Hermann, et. al., "Validation of the SCALE System for PWR Spent Fuel Isotopic
Composition Analyses,” ORNL/TM-12667, Oak Ridge National Laboratory, March 1995.

Appendix 3 A3-8


5.0 SUMMARY AND CONCLUSIONS

This study documents an evaluation of spent fuel pool (SFP) accident risk at decommissioning
plants. The study was undertaken to develop a risk-informed technical basis for reviewing
exemption requests and a regulatory framework for integrated rulemaking. The staff tried to
actively involve the public and industry representatives throughout the study. The staff held a
series of public meetings with stakeholders during and after the preparation of a preliminary
study (published in June 1999 at the request of the Nuclear Energy institute (NEI)).

The staff published a draft study in February 2000 for public comment and significant comments
were received from the public and the Advisory Committee on Reactor Safeguards (ACRS). To
address these comments the staff did further analyses and also added sensitivity studies on
evacuation timing to assess the risk significance of relaxed offsite emergency preparedness
requirements during decommissioning. The staff based its sensitivity assessment on the
guidance in Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk
Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.”
The staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet
the design and operational characteristics assumed in the risk analysis. These characteristics
are identified in the study as industry decommissioning commitments (IDCs) and staff
decommissioning assumptions (SDAs). Provisions for confirmation of these characteristics
would need to be an integral part of rulemaking.

The results of the study indicate that the risk at SFPs is low and well within the Commission's
Quantitative Health Objectives (QHOs). The risk is low because of the very low likelihood of a
zirconium fire even though the consequences from a zirconium fire could be serious. Because
of the importance of seismic events in the analysis, and the considerable uncertainty in seismic
hazard estimates, the results are presented for both the Lawrence Livermore National
Laboratory (LLNL) and the Electric Power Research Institute (EPRI) seismic hazard estimates.
In addition, to address a concern raised by the ACRS, the results also include a sensitivity to a
large ruthenium and fuel fines release fraction. The results indicate that the risk is well below
the QHOs for both the individual risk of early fatality and the individual risk of latent cancer
fatality.

The study includes use of a pool performance guideline (PPG) as an indicator of low risk at
decommissioning facilities. The recommended PPG value for events leading to uncovery of the
spent fuel was based on similarities in the consequences from a SFP zirconium fire to the
consequences from a large early release event at an operating reactor. A value equal to

the large early release frequency (LERF) criterion (1x10° per year) was recommended for the
PPG. By maintaining the frequency of events leading to uncovery of the spent fuel at
decommissioning facilities below the PPG, the risk from zirconium fires will be low and
consistent with the guidance in RG 1.174 for allowing changes to the plant licensing basis that
slightly increase risk. With one exception (the H.B. Robinson site) all Central and Eastern sites
which implement the IDCs and SDAs would be expected to meet the PPG regardless of whether
LLNL or EPRI seismic hazard estimates are assumed. The Robinson site would satisfy the
PPG if the EPRI hazard estimate is applied but not if the LLNL hazard is used. Therefore,
Western sites and Robinson would need to be considered on a site-specific basis because of
important differences in seismically induced failure potential of the SFPs.

5-1


percentage of the population evacuating from 99.5% and 95%. For cases with a small ruthenium
release, the main effect of decreasing the amount of fuel releasing fission products was a large reduction
in prompt fatalities. However, for cases with a large ruthenium release, the prompt fatalities did not
change as much, because most of the ruthenium is in the final core offload due to its one-year half-life.
With regard to the percentage of the population evacuating, the main difference between 99.5% and 95%
evacuation is in the area of early fatalities for cases with evacuation before release. In these cases, the
number of early fatalities increases by a factor of ten, because a change from 99.5% to 95% is a factor-
of-ten increase in the number of persons not evacuating.

The sensitivity calculations also showed that increasing the plume heat content resulted in reductions in
early fatalities and no change in societal dose or cancer fatalities. In addition, updating the values of the
plume-spreading parameters to those in the NUREG/CR-6244 expert elicitation results in a decrease in

early fatalities and up to a 60% increase in societal dose and cancer fatalities, because of the additional
plume spreading associated with the updated plume-spreading parameter values.

References

1. Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82, NUREG/CR-4982, July
1987

2. Code Manual for MACCS2, NUREG/CR-6613, May 1998

3. Analysis of Plume Energy from Air Oxidation in Spent Fue! Storage Pool, Sandia National
Laboratories, August 7, 2000

4. Task 7 Letter Report: Investigation of Plume Spreading Uncertainties on the Radiological
Consequences Associated with a Spent Fue! Pool Accident, Sandia National Laboratories, June 2000

-12-


runs were not available, the results for the closest corresponding calculation were used as an
approximation. The risk increase associated with the EP relaxation is the product of the event
frequency and the change in consequences, summed over all contributors.

The sensitivity of the risk increase estimates is strongly dependent on the assumptions
regarding the effectiveness of emergency evacuation in seismic events, since these events
dominate the SFP fire frequency. In NUREG-1150, evacuation in seismic events was treated
either of two ways depending on the peak ground acceleration (PGA) of the earthquake:

e for low PGA earthquakes (<0.6g), the population was assumed to evacuate however the
evacuation was assumed to start later and proceed more slowly than evacuation for
internally-initiated events. A delay time of 1.5 times the normal delay time and an
evacuation speed of 0.5 times the normal evacuation speed was assumed for this case.

e for high PGA earthquakes (>0.6g), it was assumed that there would be no effective
evacuation and that many structures would be uninhabitable. The population in the
emergency response zone was modeled as being outdoors for the first 24 hours, and
then relocating at 24 hours.

Since the SFP fire frequency is driven by seismic events with PGA several times larger than the
SSE, the reasoning that there would be no effective evacuation was adopted in developing the
baseline estimate of the risk. This is consistent with the expert opinion provided in

Attachment 2 to Appendix 2B regarding the expected level of collateral damage within the
Emergency Planning Zone given a seismic event large enough to fail the SFP. Specifically, for
ground motion levels that correspond to SFP failure in the Central and Eastern U.S., it is
expected that electrical power would be lost and more than half of the bridges and buildings
(including those housing communication systems and emergency response equipment) would
be unsafe even for temporary use within at least 10 miles of the plant. This reasoning is also
consistent with previous Commission rulings on San Onofre and Diablo Canyon in which the
Commission found that for those risk-dominant earthquakes that cause very severe damage to
both the plant and the offsite area, emergency response would have marginal benefit because
of its impairment by offsite damage.

This same reasoning is applied to the full EP and the relaxed EP cases. The net effect is that
EP, as well as relaxations in EP, do not impact the risk associated with seismic events that
result in SFP failure. A sensitivity case was also performed to explore the impact on risk
increase if the seismic event only partially degrades the emergency response, as discussed
below.

In the sensitivity case, it was assumed that evacuation would be carried out consistent with the
NUREG-1150 model for low g earthquakes if current EP requirements are maintained, i.e., the
population evacuates, but the evacuation delay time is increased by 50 percent and the time to
complete the evacuation is doubled. This is extremely optimistic given the damage to
communication and notification systems, buildings and structures, and roads that would
accompany any seismic event severe enough to fail the SFP. With no preplanning for
radiological accidents, the evacuation delay time was further increased to three times the
normal delay time.

For purposes of assigning consequences in the seismic sensitivity case, the "full EP” case was
represented by the results from the early evacuation case (i.e., evacuation is started and
completed before the release) and the "relaxed preplanning for radiological accidents" case
was represented by the results from the late evacuation case (i.e., evacuation is not started
until after the release has occurred). This maximizes the effectiveness of evacuation in the full

Appendix 4D AAD-2


floor, so that make-up can be provided even when the environment is uninhabitable
because of steam and/or high radiation (NEI commitment no.8).

. Repair crew is different than onsite operators.

. Repair crew will focus recovery efforts only on one pump.

. On average, it takes 10 hours to repair a pump if it fails to start and run.

e It takes 16 hours to contact maintenance personnel, make a diagnosis, and get new parts.

e Both firewater pumps are located in a separate structure or protected from the potential
harsh environment in case of pool bulk boiling.

. Maintenance is performed per schedule on diesel and electric firewater pumps to maintain
operable status.

. Operators have received formal training on relevant procedures.

4.3.4.3 Quantification

Human Error Probabilities
The fault tree LP!I-OMK-U includes five human failure events and LPI-OMK-L has three.

Two events are common. HEP-RECG-FWSTART represents the failure of the operator to
recognize the need to initiate firewater as an inventory makeup system, given that a loss of fuel
pool cooling has been recognized. This event was quantified using the SPAR HRA technique.
The assumptions included expansive time (> 24 hours), a high level of stress, diagnostic type
procedures, good ergonomic interface, and good quality of work process.

HEP-FW-START represents failure to start either the electric or diesel firewater pump
(depending upon availability) within 88 hours after the onset of bulk boiling, given that the
decision to start a firewater pump was made. No difficult valve alignment is required, but the
operator may have to run hoses to designated valve stations. This event was quantified using
the SPAR HRA technique. The PSFs included expansive time (> 50 times the required time),
high stress, highly complex task because of the multiple steps, its non-routine nature, quality
procedures available, as well as good ergonomics including equipment and tools matched to
procedure, and finally a crew who had executed these tasks before, conversant with the
procedures and one another.

HEP-FW-REP-NODEP represents the failure of the repair crew to repair a firewater pump for
the scenario where power is not recovered. Note that it has been assumed that since power is
not recovered, the repair crew did not make any attempt to repair the SFPC system, and
therefore no dependency was modeled in the failure to repair the firewater system. Assuming
that it takes another 16 hours before technical help and parts arrive, then the operator has

72 hours (88 hours less 16 hours) to repair the pump. Assuming a 10-hour mean time to repair,

Appendix 2A A2A-38


Assuming that it takes 16 hours before technical help and parts arrive, then the operators have
72 hours (88 hours less 16 hours) to repair the pump. Assuming a 10-hour mean time to repair,
the probability of failure to repair the pump would be Exp [-(1/10)x72] = 1.0E-3.

Hardware Failure Probabilities

Basic event FP-DGPUMP-FTF represents the failure of the diesel-driven firewater pump. The
pump may be required to run 8 to 10 hours at the most (250 gpm capacity), given that the water
inventory drops by 20 ft (i.e., 3 ft from the top of the fuel). A failure probability of 1.8E-1 for
failure to start and run for the diesel driven pump is used from INEL-96/0334 (Ref. 15).

4.2.5.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-RECG-FWSTART 2.0E-5
HEP-FW-START 1.0E-5
HEP-FW-REP-NODEP 1.0E-3
EP-DGPUMP-FTF 1.8E-1

4.2.6 Top Event OFD - Operator Recovery Using Offsite Sources

4.2.6.1 Event Description and Timing

Given the failure of recovery actions using onsite sources, this event accounts for recovery of

coolant makeup using offsite sources. Adequate time is available for this action, provided that

the operators recognize that recovery of cooling using onsite sources will not be successful, and

that offsite sources are the only viable alternatives. This top event is quantified using fault tree

FIR-OFD. This event is represented by a basic event HEP-INV-OFFSITE.

4.2.6.2 Relevant Assumptions

° The operators have 88 hours to provide make-up and inventory cooling.

. Procedures and training are in place that ensure that offsite resources can be brought to
bear (NEI commitment no. 2 and 4), and that preparation for this contingency is made
when it is realized that it may be necessary to supplement the pool make-up.

. Procedures explicitly state that if the water level drops below a certain level (e.g., 15 ft
below normal level) operator must initiate recovery using offsite sources.

¢ Operators have received formal training in the procedures.

° Offsite resources are familiar with the facility.

Appendix 2A A2A-31


assured, a zirconium fire cannot be precluded, although the likelihood may be reduced by
accident management measures.

Figure 2.1 plots the heatup time air-cooled PWR and BWR fuel take to heat up from 30 °C to
900 °C versus time since reactor shutdown. The figure shows that after 4 years, PWR fuel
could reach the point of fission product release in about 24 hours. Figure 2.2 shows the timing
of the event by comparing the air-cooled calculations to an adiabatic heatup calculation for PWR
fuel with a burnup of 60 GWD/MTU. The figure indicates an unrealistic result that until 2 years
have passed the air-cooled heatup rates are faster than the adiabatic heatup rates. This is
because the air-cooled case includes heat addition from oxidation while the adiabatic case does
not. In the early years after shutdown, the additional heat source from oxidation at higher
temperatures is high enough to offset any benefit from air cooling. This result is discussed
further in Appendix 1A. The results using obstructed airflow (adiabatic heatup) show that at

5 years after shutdown, the release of fission products may occur approximately 24 hours after
the accident.

In summary, 60 days after reactor shutdown for boildown type events, there is considerable time
(>100 hours) to take action to preclude a fission product release or zirconium fire before
uncovering the top of the fuel. However, if the fuel is uncovered, heatup to the zirconium ignition
temperature during the first years after shutdown would take less than 10 hours even with
unobstructed air flow. After 5 years, the heatup would take at least 24 hours even with
obstructed air flow cases. Therefore, a zirconium fire would still be possible after 5 years for
cases involving obstructed air flow and unsuccessful accident management measures. These
results and how they affect SFP risk and decommissioning regulations are discussed in
Sections 3 and 4 of this study.

zirconium fire would not occur because of a lack of oxygen. Blockage of the air flow around the
fuel could be caused by collapsed structures and/or a partial draindown of the SFP coolant or by
reconfiguration of the fuel assemblies during a seismic event or heavy load drop. A loss of SFP
building ventilation could also preclude or inhibit effective cooling. As discussed in Appendix 1A,
air flow blockage without any recovery actions could result in a near-adiabatic fuel heatup and a
zirconium fire even after 5 years.

2-2


consequence assessment model for Surry. The results of the Base Case are shown in Table
A4-7.

Table A4-7 Mean Consequences for the Base Case

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 1.75 47,700 2,460
0-500 1.75 571,000 25,800

90 days 0-100 1.49 46,300 2,390
0-500 1.49 586,000 26,400

1 year 0-100 1.01 45,400 2,320
0-500 1.01 595,000 26,800

Table A4-7 shows the offsite consequences for a severe SFP accident at 30 days, 90 days,
and 1 year following final reactor shutdown. The decay times for fuel transferred to the pool
during the 11" refueling outage were 30 days, 90 days, and 1 year, respectively. The decay
times for spent fuel in the pool from earlier refueling outages were much longer and were
accounted for in the inventories used in this analysis.

These results in Table A4-7 show virtually no change in long-term offsite consequences (i.e.,
societal dose and cancer fatalities) as a function of decay time, because they are controlled by
inventories of radionuclides with long half-lives and relocation assumptions. However, these
results also show about a factor-of-two reduction in the short-term consequences (i.e., prompt
fatalities) from 30 days to 1 year of decay. (All of the prompt fatalities occur within 10 miles of
the site.) As a rough check on the prompt fatality results, the change in decay power was
evaluated for an operating reactor shut down for 30 days and for 1 year. The decay power
decreased by about a factor of three. This is consistent with a factor-of-two decrease in prompt
fatalities. The factor-of-three decrease in decay power by radioactive decay will also increase
the time it takes to heat up the spent fuel, which provides additional time to take action to
mitigate the accident.

The results of Case 1, which used a lower evacuation percentage than the Base Case, are
identical to the results of the Base Case shown in Table A4-7. Case 1 used an evacuation
percentage of 95%, while the Base Case used an evacuation percentage of 99.5%. Although it
might be expected to see an increase in prompt fatalities from reducing the evacuation
percentage, no such increase was observed. This is because of the assumption that the
telease ends at 1.2 hours, while the evacuation does not begin until 2 hours.

Case 2, shown in Table A4-8, used a radionuclide inventory that consisted of 11 batches of
spent fuel, but did not include the remaining two-thirds of the core in the vessel. This was done
to facilitate comparison of the consequence results with the results of the analyses in
NUREG/CR-4982 and NUREG/CR-6451. This also allowed examination of the relative
contribution of the short-lived radionuclides to consequences. Because the length of time

Appendix 4 A4-9



Table 4.2 Basic Event Summary for the Internal Fire Event Tree

Basic Event
Basic Event Name Description Probability
Internal fire initiating event
IE-FIRE 3.0E-3
Operators fail to respond to a signal
HEP-DIAG-ALARM indication in the control room 3.0E-4
HEP-RES-FIRE Operators fail to suppress fire 2.5E-1
Operators fail to observe the loss of
HEP-WLKDWN-LSFPC. cooling in walk-downs (independent 1.0E-5
case)
¥ Operators fail to diagnoses need to
HEP-RECG-FWSTART start the firewater system 2.0E-5
HEP-FW-START Operators fail to start firewater pump 1.0E-5
and provide alignment
Repair crew fails to repair firewater
HEP-FW-REP-NODEP system 1.0E-3
HEP-INV-OFFSITE Operators fail to provide alternate 5.0E-2
sources of cooling from offsite
FP-DGPUMP-FTF Failure of firewater pump system 0.18
SFP-FIRE-LOA Electrical faults causing loss of alarms 2.0E-3
SFP-FIRE-DETECT Failure of fire detectors 5.0E-3

4.3 Plant-centered and Grid-related Loss of Offsite Power Event Tree

This event tree represents the loss of SFP cooling resulting from a loss of offsite power from
plant-centered and grid-related events. Until offsite power is recovered, the electrical pumps
would be unavailable, and only the diesel fire pump would be available to provide make-up.

Figure 4.3 shows the Plant-centered and Grid-related Loss of Offsite Power (LOSP) event tree
sequence progression.

4.3.1 Initiating Event LP1 - Plant-centered and Grid-related Loss of Offsite Power

4.3.1.1 Event Description

Initiating event IE-LP1 represents plant-centered and grid-related losses of offsite power.
Plant-centered events typically involve hardware failures, design deficiencies, human errors (in

Appendix 2A A2A-33


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3-41


4.1.3 Safety Margins ... 0.2.20... ce ee eee! 4-8
4.1.4 Implementation and Monitoring Program 249
4.2 Implications for Regulatory Requirements for Emergency Preparedness, Security,

and Insurance 2.0.0.0... cece cece eee tenet e eee ener  AN3

4.2.1 Emergency Preparedness ...... 4-13

4.2.2 Security .................-. 4-14

4.2.3 Insurance 2.0.0... eee eet nt e eter eed 4-15

5.0 SUMMARY AND CONCLUSIONS ..... 00.06.6622 5-1

6.0 REFERENCES ............. 0.200 c cece cece een teen teens 6-1

7.0 ACRONYMS ......0. 2.0.0. cece ccc e cece nett e nee e es DA

List of Figures

ES-1 Individual Early Fatality Risk Within 1 Mile... 2... 2.2... 6. eee eee eee eee xii

ES-2 Individual Latent Cancer Fatality Risk Within 10 Miles . . Xi

Figure 2.1 Heatup Time From 30 °C to 900 °C...............-.--. 22-3

Figure 2.2 PWR Heatup Times for Air Cooling and Adiabatic Heatup - +. 2-3

Figure 3.1 Assumed Spent Fuel Pool Cooling System .......-. 6.65.66. sees eee eee 3-4
Figure 3.2 Frequency of Spent Fuel Pool Seismically Induced Failure Based on LLNL

Estimates and an HCLPF of 1.2 g Peak Spectral Acceleration ............. 3-21

Figure 3.3. Frequency of Spent Fuel Pool Seismically Induced Failure Based on EPRI
Estimates and an HCLPF of 1.2 g Peak Spectral Acceleration..............
Figure 3.7-1 Early Fatality Consequences for Spent Fuel Pool Source Terms ..
Figure 3.7-2 Societal Dose Consequences for Spent Fuel Pool Source Terms
Figure 3.7-3 Spent Fuel Pool Early Fatality Risk .............--.--------5- see
Figure 3.7-4 Spent Fuel Pool Societal Risk ........... 2.0020 eee e rece eee
Figure 3.7-5 Sensitivity of Early Fatality Risk to Emergency Planning — Cask Drop

Event 20.0... eee teen ett e ete n eens 3-43

Figure 3.7-6 Sensitivity of Societal Risk to Emergency Planning — Cask Drop Event .. . 3-44

Figure 3.7-7 _ Individual Early Fatality Risk Within 1 Mile ...............---.---+-- 3-47

Figure 3.7-8 Individual Latent Cancer Fatality Risk Within 10 Miles ..............-- 3-48
List of Tables

Table 2.1 Time to Heatup and Boiloff SFP Inventory Down to 3 Feet Above Top of Fuel

(60 GWD/MTU) .. 2.22 eee eee nee
Table 3.1 Spent Fuel Pool Cooling Risk Analysis — Frequency of Fuel Uncovery ....
Table 3.2 Spent Fuel Pool Cooling Risk Analysis — Frequency Partition for Air Flow . . 3-10
Table 3.7-1 Consequences of an SFP Accident With a High Ruthenium Source Term .. 3-29
Table 3.7-2 Consequences of an SFP Accident With a Low Ruthenium Source Term .. 3-30

Table 3.7-3 Frequency of Boil Down Events Leading to Spent Fuel Uncovery ........-. 3-35
Table 3.7-4 Mean Frequency of Rapid Draindown Due to Seismic Events............- 3-37
Table 3.7-5 Frequency of Rapid Draindown Spent Fuel Uncovery Due to Nonseismic

Events 22... 6... e nent e eee 3-38


ATTACHMENT A

FAULT TREES USED IN THE RISK ANALYSIS


SPAR HRA Human Error Worksheet (Page 3 of 3)

Plant: Initiating Event: Sequence Number: Basic Event Code: __

PART Ill. CALCULATE THE TASK FAILURE PROBABILITY WITHOUT FORMAL
DEPENDENCE (Pwion)

Calculate the Task Failure Probability Without Formal Dependence (P,,,,,) by adding the Diagnosis
Failure Probability (from Part I, p.1) and the Action Failure Probability (from Part II, p. 2).

If all PSFs are
nominal, then
Diagnosis Failure Probability: Diagnosis Failure Probability:
1E-2
Action Failure Probability: + Action Failure Probability:
+1E-3

Task Failure Without
Formal Dependence (Pyjog)= Prwied) = 11E-2


Table 1

Seismic Hazard Estimates for Peak Spectral Acceleration for PSA

From Refs. 8 and 9 (LLNL 93 Results

Peak Spectral Acceleration PSA
(cm/sec.?
Exceedance
Frequency Vermont Yankee Robinson
H
1x10* 93 232
5x10+ 151 369
2x10* 246 676
1x104 354 991
5x10% 501 1349
2x10° 759 2054
ix10* 1058 2801
5x10° 1396 3915
2x10* 1884 6096
1x10* 2308 8522
5x107 2661 -
2x107 3330 -
1x107 3802 -
5x10* 4266 -
2x10* 5248 -
* By Interpolation
** By Extrapolation
Table 2

**
et
ae
eK
ak

Comparison of Seismic Risk Estimated by Various Approaches

Cucuer = 1.2g PSA, B=0.4

Computed Seismic Risk P-
(to be multiplied by 10%)
Site Ref. 1 Method Ref. 10 Method Rigorous
Egn. (3) Steps 1 through 3_|_ Convolution
Vermont Yankee 0.38 1.07 0.89
Robinson 3.7 13.6 13.0

A2B-21



4.5.5.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-LEAK-ISO 1.3E-3
HEP-MKUP-START-E 2.5E-4
SFP-REGMKUP-F 5.0E-2

4.5.6 Top Event OIL - Operator Initiates SFP Makeup System
4.5.6.1 Event Description and Timing

This top event represents the failure to initiate the SFP makeup system in time to prevent loss of
SFP cooling, for a small leak. This top event is represented by the fault trees LOI-OIL-U and
LOI-OIL-L, which include contributions from operator error and hardware failure. The leak is
small enough that isolation is not required for success. If the operators respond to the initiator
early (i.e., CRA is successful), they would have more than 8 hours to terminate the event using
the SFP makeup system before the water level drops below the SFP suction level. If operators
respond late (i.e., IND success), it is assumed that they would have on the order of 4 hours,
based on the leak initiating at the start of one shift and the walkdown taking place at shift
turnover.

4.5.6.2 Relevant Assumptions

. There are procedures to guide the operators in how to deal with loss of inventory, and the
operators are trained in their use (NEI commitment no. 2).

¢ The manipulations required to start the makeup system can be achieved in less than 10
minutes.

4.5.6.3 Quantification
Human Error Probabilities

In the case of an early response, the operator would have more than 8 hours available to
establish SFP make-up and the failure is represented by the basic event HEP-MKUP-START
(see fault tree L OI-OIL-U). In the case of a late response, the operator is assumed to have

4 hours available to establish SFP make-up and is represented by the basic event HEP-MKUP-
START-E (see fault tree L OI-OIL-L). Success is defined as the operator starting the makeup
pump and performing valve manipulation as needed.

SPAR HRA worksheets were used to quantify each of these errors. For HEP-MKUP-START it
was assumed that the 8 hour time window will allow more than 50 times the time required to
complete this task, the operators are under high stress, are highly trained, have equipment that
is well labeled and matched to a procedure, and the crew has effective interactions in a quality
facility. For HEP-MKUP-START-E, the time available is not as extensive, and is considered
nominal, all other PSFs being equal.

Appendix 2A A2A-59


place provide adequate margins for accident prevention. Additionally, the presence of soluble
boron or Boraflex provides additional assurance of margin with respect to shutdown reactivity.

The risk results provided in Section 3.6.3 show that a typical site that conforms with the IDCs
and SDAs would meet the Commission’s QHOs by one to two orders of magnitude, depending
on assumptions about the SFP source term and seismic hazard frequency. The risk
comparisons provided in Appendix 4C show that SFP facilities maintained at or below the
recommended PPG of 1x10° per year would continue to meet the QHOs for even the most
severe source term.

The estimated risk increases associated with the EP relaxations are also well below the
allowable increases developed from the RG 1.174 LERF criteria. As discussed in Section 4.2.1
and Appendix 4D, the increases in risk from the EP relaxation would be about a factor of

10 below the maximum allowable increases developed from RG 1.174. Since the SFP fire
frequency assumed in the RG 1.174 comparisons is about a factor of 4 lower than the PPG of
1x10° per year, a plant operating nominally at the PPG would have a smaller margin to the
allowable risk limits for the reference plant but would still be at or below the limits.

The results of a sensitivity case in Appendix 4D indicate that even with assumptions that
maximize the effectiveness of EP in seismic events, the change in risk associated with
relaxation of the requirements for radiological preplanning is still relatively small. The increases
in early fatalities and individual early fatality risk remain below the maximum allowable for each
risk measure. Population dose and individual latent cancer fatality risk are about a factor of 2
higher than the allowable value inferred from RG 1.174. This increase in individual latent cancer
risk represents about 9 percent of the QHO; thus, considerable margin to the QHO would still
remain.

The evacuation effectiveness assumed for “full EP” in the sensitivity case is unrealistic for high
ground motion earthquakes, and the risk increase associated with the EP relaxations is
expected to be closer to the baseline value. Also, the risk reduction estimates are based on the
LLNL seismic hazard frequencies and the high ruthenium source term, and would be
substantially lower if either the EPRI seismic hazard frequencies or the low ruthenium source
term were used. The above comparisons are based on the risk levels 1 year after shutdown but
would also be valid several months following shutdown. Use of either the EPR! seismic hazard
frequencies or the low ruthenium source term would reduce each of the risk measures by about
a factor of 10, to values well below the RG 1.174 guidelines and the QHOs. The risk impact will
decrease even further in later years due to reduced consequences as fission products decay.

The study concludes that relaxation of certain EP requirements can be considered for
decommissioning plants in which conformance with the IDCs and SDAs provides reasonable
assurance that sufficient margins to the safety goals will be maintained.

4.1.4 Implementation and Monitoring Program

RG 1.174 states that the impact of the proposed change should be monitored using
performance measurement strategies. RG 1.174 further states that an implementation and
monitoring plan should be developed to ensure that the engineering evaluation conducted to

4-9


Because the uncovery frequency is already very low (on the order of 1 in 1 million per year) both
absolutely and relative to other initiators, and because the quantification of human reliability
analysis values for such extended periods of recovery is beyond the state-of-the-art, the staff did
not attempt to recalculate the expected uncovery frequency. For 2-year-old, 5-year-old, and
10-year-old fuel, much longer periods are available than at 1 year (see Table 2.1).

A careful and thorough adherence to IDCs #2, #5, #8, and #10 is crucial to establishing and
maintaining the low frequency. In addition, however, the assumption that walkdowns are
performed on a regular basis (once per shift) is important to compensate for potential failures of
the instrumentation monitoring the status of the pool. The analysis has also assumed that the
procedures and/or training give explicit guidance on the capability of the fuel pool makeup
system and on when it becomes essential to supplement with alternative higher volume sources.
The analysis also assumes that the procedures and training give sufficiently clear guidance on
early preparation for using the alternative makeup sources.

There have been two recent events involving a loss of cooling at SFPs. The first, at Browns
Ferry Unit 3 occurring in December 1998, involved a temperature increase of approximately

25 °F over a 2-day period. This event, caused by the short cycling of cooling water through a
stuck-open check valve, was not detected by the control room indicators because of a design
flaw in the indicators. In the second event, at the Duane Arnold Unit 1 in January 2000, the SFP
temperature increased by 40 to 50 °F. The incident, which was undetected for approximately
2% days, was caused by operator failure to restore the SFP cooling system heat sink after
maintenance activities. The plant had no alarm for high fuel pool temperature, although there
are temperature indicators in the control room. Since the conditional probability of fuel
uncovery is low given a loss of cooling initiating event, the addition of these two recent events to
the database will not affect the conclusion that the risk from these events is low. However, the
recent events illustrate the importance of industry commitments, particularly IDC #5 which
requires temperature instrumentation and alarms in the control room. In addition, the staff
assumptions that walkdowns are performed on a regular basis (once per shift), with the fuel
handler documenting the observations in a log, and the assumption that control room
instrumentation that monitors SFP temperature and water level directly measures the
parameters involved are important for keeping the risk low, since the walkdowns compensate for
potential failures of the control room instrumentation and direct measurement precludes failures
such as occurred at Browns Ferry.

Even with the above referenced industry commitments, the additional need for walkdowns to be
performed at least once per shift and the specific need for direct indication of level and
temperature had to be assumed in order to arrive at the low accident frequency calculated for
this scenario. These additional assumptions are identified by the staff as staff decommissioning
assumptions (SDAs) #2 and #3. SDA #2 assumes the existence of explicit procedures and fuel
handler training to provide guidance on the capability and availability of inventory makeup
sources and the time available to utilize these sources.

SDA #2 Walkdowns of SFP systems are performed at least once per shift by the fuel handlers.
Procedures are in place to give the fuel handlers guidance on the capability and
availability of onsite and offsite inventory makeup sources and on the time available to
utilize these sources for various loss of cooling or inventory events.

3-13


most SFP accident scenarios allows for reasonable human recovery actions to respond to
system failures, and provides sufficient time to allow for the implementation of protective actions.

The staff's risk assessment demonstrates that the risk from a decommissioning plant SFP
accident is small if IDCs and additional SDAs are implemented as assumed in the risk study.
Due to the different nature of an SFP accident versus an accident in an operating reactor with
respect to system design capability and event timing, the defense-in-depth function of reactor
containment is not required. However, the staff has found that defense-in-depth in the form of
accident prevention measures and an appropriate level of emergency planning can limit risk and
provide dose savings for as long as a zirconium fire is possible.

Defense-in-depth for accident prevention and mitigation is provided by licensee conformance
with the IDCs and SDAs, as discussed previously. Defense-in-depth for consequence mitigation
should continue to be provided by retaining requirements for an appropriate level of EP in
consideration of the amount of time available before fission product release in specific events.

For the purpose of the analysis in this study, when referring to relaxation of offsite EP, the study
assumed conditions that would be similar to those at sites in decommissioning that have already
received exemptions from some EP requirements. For instance, licensees may no longer be
required: to have a formalized emergency planning zone (EPZ); to have an emergency
operations facility (EOF), technical support center (TSC), or operations support center (OSC); to
promptly notify the public using a siren system, tone alert radios, or National Weather Service
radios; or to conduct biennial full-participation exercises. The analyses in the study were
simplified to focus on conditions which assumed evacuation occurred either early or late.

It is understood that EP involves more than just evacuation considerations. In the analysis of
the study, it was assumed that the decommissioning licensee would still be required to notify
offsite authorities, characterize the releases, and make protective action recommendations;
have a means of notifying offsite organizations and providing information to the public; and hold
onsite biennial exercises and semiannual drills.

The assessments conducted for this study show that, 60 days after final shutdown, recovery and
mitigation times of more than 100 hours are-available before release occurs, except for the most
severe events. These times appear to be sufficient to permit offsite protective actions to be
implemented on an ad hoc basis, if necessary, without the full compliment of regulatory
requirements associated with operating reactors. The staff notes that potential relaxation of EP
requirements for decommissioning plants could be phased in such that the relaxation would not
result in an immediate lapse of all offsite emergency response capabilities following final
shutdown, but would more likely result in early elimination of some capabilities (e.g., sirens) and
more gradual relaxation of certain other capabilities (e.g., pre-planning of evacuations and
communications), with a transition towards longer ad hoc response times over several years
due to such factors as attrition of experienced personnel. Shortly after final shutdown, when
SFP heatup rates and risks are greatest, response capabilities are expected to be largely intact
and comparable to those for full EP. These capabilities could be expected to diminish over time,
resulting in longer ad hoc response times. However, continued fission product decay in the
spent fuel will result in longer times to release, providing additional time during which emergency
response measures could be implemented.

47


Table 4.3 Basic Event Summary for the Plant-centered and Grid-related Loss of Offsite

Power Event Tree
Basic Event Name Description Probability
IE-LP1 Loss of offsite power because of 8.0E-2
plant-centered or grid-related causes
REC-OSP-PC Recovery of offsite power within 24 hours 1.0E-2
HEP-DIAG-SFPLP1 Operators fail to diagnose loss of SFP 1.0E-6

cooling because of loss of offsite power

HEP-FW-REP-DEPEN | Repair crew fails to repair firewater system 5.0E-2
- dependent case

HEP-SFP-STR-LP1 Operators fail to restart and align the SFP 5.0E-6
cooling system once power is recovered

HEP-RECG-FWSTART | Operators fail to diagnose need to start the 2.0E-5
firewater system

HEP-RECG-DEPEN Operators fail to recognize need to cool 5.0E-02
pool given prior failure

HEP-FW-START Operators fail to start firewater pump and 1.0E-5
provide alignment

HEP-FW-REP-NODEP | Repair crew fails to repair firewater system 1.0E-3

SPC-PMP-CCF SFP cooling pumps - common cause 5.9E-4
failure

SPC-PMP-FTF-1 SFP cooling pump 1 fails to start and run 3.9E-3

SPC-PMP-FTF-2 SFP cooling pump 2 fails to start and run 3.9E-3

FP-2PUMPS-FTF Failure of firewater pump system 6.7E-4

FP-DGPUMP-FTF Failure of the diesel-driven firewater pump 1.8E-1
Heat exchanger discharge check valves- 1.9E-5

SPC-CKV-CCF-H CCF

SPC-CKV-CCF-M SFP cooling pump discharge check 3.2E-5
valves-CCF

SPC-HTX-CCF SFP heat exchangers - CCF 1.9E-5

SPC-HTX-FTR SFP heat exchanger cooling system fails 2.4E-4

SPC-HTX-PLG Heat exchanger plugs 2.2E-5

Appendix 2A A2A-42


The individual early fatality risk for an SFP accident is about one to two orders of magnitude
lower than the Commission's safety goal, depending on assumptions about the SFP accident
source term and seismic hazard. For the upper curve (corresponding to use of the mean of the
LLNL seismic hazard estimates and the high ruthenium source term), the risks are about a
decade lower than the safety goal. For the lower curve (corresponding to use of the mean of the
EPRI seismic hazard estimates and the low ruthenium source term) the risks are about

2 decades lower than the safety goal. The individual early fatality risk for an SFP accident
decreases with time and is about a factor of 5 lower at 5 years following shutdown (relative to
the value at 30 days).

Similarly, the individual latent cancer fatality risk for an SFP accident is about one to two orders
of magnitude lower than the Commission's safety goal, depending on assumptions about the
SFP accident source term and seismic hazard. For the upper curve (corresponding to use of
the mean of the LLNL seismic hazard estimates and the high ruthenium source term), the risks
are about a decade lower than the safety goal. For the lower curve (corresponding to use of the
mean of the EPRI seismic hazard estimates and the low ruthenium source term), the risks are
about 2 decades lower than the safety goal. The individual latent cancer fatality risk for an SFP
accident are not substantially reduced with time because of the slower decay of fission products
and the interdiction modeling assumptions that drive long-term doses.

Changes to EP requirements, as modeled, do not substantially impact the margin between SFP.
risk and the safety goals because of the low frequency of events for which EP would be
effective.


to include access to portable pumps and emergency power. IDC #7 and IDC #9 commit the
industry to implement procedures or administrative controls to reduce the likelinood of rapid
draindown events. SDA #2 calls for procedures to be developed that will provide guidance on
the availability of on-site and off-site inventory make-up sources and time available to initiate
these sources. In addition, the industry has committed in IDC #10 to perform routine testing of
the alternative spent fuel pool make-up system components and to have procedural controls on
equipment out of service to increase confidence that components will be available. The two
accidents that could lead to very rapid draining of the SFP are extremely large seismic events
and heavy load drops. IDC #1 and SDA #5 address heavy load drop concerns. SDA #6 calls
for each decommissioning plant to successfully complete the seismic checklist provided in
Appendix 2b to this report. Implementation of these commitments and assumptions will help
assure the frequency of a zirconium fire remains within the assumptions of the analysis.

Comment #16: Is station blackout at a decommissioning site acceptable to the staff?

Response: As with an operating reactor, the staff recognizes that there is some small annual
probability that a station blackout will occur at a decommissioning site. For both an operating
and decommissioning plant, there is a need to recover from a station blackout. However,
unlike an operating reactor, decommissioning SFPs can go without electrical power for a long
period of time (days to weeks) and not suffer safety consequences. This is due to the inherent
margin provided by the large volume of water sitting above the spent fuel in the pool. It takes a
long time to heat this water up to boiling and then to continue to boil it off until fuel is uncovered
(almost a week at one year after the last fuel was irradiated in the reactor). IDC #2 commits the
industry to develop procedures and train personnel to ensure that on-site and off-site resources
can be brought to bear during an event. IDC #3 calls for communication systems to be set up
between the SFP site and off-site resources that can survive severe weather and seismic
events, which can cause a station blackout.

Comment #17: The risk assessment should take into account changes in local aircraft traffic
when evaluating the probability and consequences from aircraft crashing into SFPs.

Response: The risk from aircraft crashes is small, and even large increases in traffic should
not make aircraft crashes.a dominant contributor to risk. A decommissioning plant will continue
to be governed by 10 CFR Part 50 for the evaluation of hazards as discussed in Standard
Review Plan 2.2.3, “Evaluation of Potential Accidents,” including accidents involving nearby
industrial, military, and transportation facilities.

The frequency of an aircraft crash leading to an accident in a spent fuel pool was estimated in
the report to be in the range of 4.3x10* to 9.6x10"? per year where damage to the pool was
significant enough that it resulted in a rapid loss of water from the pool (See Section 3 and
Appendix 2d). The mean value was estimated to be 2.9x10° per year. These values are a
small fraction of the overall risk of uncovering the spent fuel in the poo! at a decommissioned
plant. An aircraft crash could also result in damage to a SFP support system. The estimated
range of striking a support system was estimated to be in the range of 1.0x10° to 1.0x10° per
year, with a mean value of 7.0x10” per year, without consideration of recovery actions. These
values are also a small fraction of the estimated frequencies for the loss of cooling initiator
(3.0x10° per year), the internal fire initiator (3.0x10° per year), or the loss of inventory initiator
(1.0x10°% per year).

Aircraft traffic and accident data were reviewed by the staff (Ref: “Data Development Technical

Appendix 6 A6-6


IDC #4

IDC #5

IDC #6

IDC #7

IDC #8

IDC #9

IDC #10

An offsite resource plan will be developed which will include access to portable pumps
and emergency power to supplement onsite resources. The plan would principally
identify organizations or suppliers where offsite resources could be obtained in a
timely manner.

Spent fuel pool instrumentation will include readouts and alarms in the control room
(or where personnel are stationed) for spent fuel pool temperature, water level, and
area radiation levels.

Spent fuel pool seals that could cause leakage leading to fuel uncovery in the event of
seal failure shall be self limiting to leakage or otherwise engineered so that drainage
cannot occur.

Procedures or administrative controls to reduce the likelihood of rapid draindown
events will include (1) prohibitions on the use of pumps that lack adequate siphon
protection or (2) controls for pump suction and discharge points. The functionality of
anti-siphon devices will be periodically verified.

An onsite restoration plan will be in place to provide repair of the spent fuel pool
cooling systems or to provide access for makeup water to the spent fuel pool. The
plan will provide for remote alignment of the makeup source to the spent fuel pool
without requiring entry to the refuel floor.

Procedures will be in place to control spent fuel pool operations that have the potential
to rapidly decrease spent fuel pool inventory. These administrative controls may
require additional operations or management review, management physical presence
for designated operations or administrative limitations such as restrictions on heavy
load movements.

Routine testing of the alternative fuel pool makeup system components will be
performed and administrative controls for equipment out of service will be
implemented to provide added assurance that the components would be available, if
needed.

Additional important operational and design assumptions made by the staff in the risk estimates
developed in this study are designated as SDAs and are discussed in later sections of this

study.

3.3 Estimated Frequencies of Spent Fuel Uncovery and Assumptions That Influence the
Results

Based upon the above design and operational features, IDCs, technical comments from
stakeholders, and the input from the INEEL technical review, the staff's SFP risk model was

updated.

The updates have improved the estimated frequency calculations, but have not

changed the need for the industry commitments or staff decommissioning assumptions.
Absolute values of some sequences have decreased, but the overall insights from the risk
assessment remain the same.

3-6


3.5.2 Aircraft Crashes

The staff evaluated the likelihood that an aircraft crashing into a nuclear power piant site would
seriously damage the spent fuel pool or its support systems (details are in Appendix 2D). The
generic data provided in DOE-STD-3014-96 (Ref. 6) was used to assess the likelihood of an
aircraft crash into or near a decommissioning spent fuel pool. Aircraft damage can affect the
structural integrity of the spent fuel pool or the availability of nearby support systems, such as
power supplies, heat exchangers, or water makeup sources, and may also affect recovery
actions. There are two approaches to evaluating the likelinood of an aircraft crash into a
structure. The first is the point target model, which uses the area (length times width) of the
target to determine the likelihood that an aircraft will strike the target. The aircraft itself does not
have real dimensions in this model. In the second approach, the DOE model modifies the point
target approach to account for the wing span and the skidding of the aircraft after it hits the
ground by including the additional area the aircraft could cover. The DOE model also takes into
account the piane’s glide path by introducing the height of the structure into the equation, which
effectively increases the area of the target.

In estimating the frequency of catastrophic PWR spent fuel poo! damage from an aircraft crash
{i.e., the pool is so damaged that it rapidly drains and cannot be refilled from either onsite or
Offsite resources), the staff uses the point target area model and assumes a direct hit on a

100 x 50 foot spent fuel pool. Based on studies in NUREG/CR-5042, “Evaluation of External
Hazards to Nuclear Power Plants in the United States,” it is estimated that 1 of 2 aircrafts are
large enough to penetrate a 5-foot-thick reinforced concrete wall. The conditional probability
that a large aircraft crash will penetrate a 5-foot-thick reinforced concrete wall is taken as

0.45 (interpolated from NUREG/CR-5042). It is further estimated that 1 of 2 crashes damage
the spent fuel pool enough to uncover the stored fuel (for example, 50 percent of the time the
location of the damage is above the height of the stored fuel). The estimated range of
catastrophic damage to the spent fuel pool resulting in uncovery of the spent fuel is 1.3x10TM" to
6.0x10* per year. The mean value is estimated to be 4.1x10° per year. The frequency of
catastrophic BWR spent fuel pool damage resulting from a direct hit by a large aircraft is
estimated to be the same as for a PWR. Mark-! and Mark-ll secondary containments generally
do not appear to have any significant structures that might reduce the likelihood of aircraft
penetration, although a crash into 1 of 4 sides of a BWR secondary containment may be less
likely to penetrate because other structures are in the way of the aircraft. Mark-Ill secondary
containments may reduce the likelihood of penetration somewhat, since the spent fuel pool may
be protected on one side by additional structures. If instead of a direct hit, the aircraft skids into
the pool or a wing clips the pool, catastrophic damage may not occur. The staff estimates that
skidding aircraft are negligible contributors to the frequency of fuel uncovery resulting from
catastrophic damage to the pool because skidding decreases the impact velocity. The
estimated frequencies of aircraft-induced catastrophic spent fuel pool failure are bounded by
other initiators.

The staff estimated the frequency of significant damage to spent fuel pool support systems (e.g.,
power supply, heat exchanger, makeup water supply) for three different situations. The first
case is based on the DOE model including the glide path and the wing and skid area and
assumes a structure 400 x 200 x 30 feet (i.e., the large building housing the support systems)
with a conditional probability of 0.01 that one of these systems is hit (the critical system

3-23


4.5.2 Top Event NLL - Loss Exceeds Normal Makeup Capacity

4.5.2.1 Event Description and Timing

This phenomenological event divides the losses of inventory into two categories: those for
which the leak size exceeds the capacity of the SFP make-up and therefore require isolation of
the leak, and those for which the SFP makeup system's capacity is sufficient to prevent fuel
uncovery without isolation of the leak.

4.5.2.2 Relevant Assumptions

« Inthe case of a large leak, a leak rate is assumed to be twice the capacity of the SFP
makeup system, i.e., 60 gpm. Although a range of leak rates is possible, the larger leak
rates are postulated to be from failures in gates, seals, or from large siphoning events, and
NEI commitments 6 and 7 will go a considerable way toward minimizing these events.

* The small leak is assumed for analysis purposes to be at the limit of the makeup system
capacity, i.e., 30 gpm.

4.5.2.3 Quantification
Non-HEP Probabilities

This top event is quantified by a single basic event, LOI-LGLK. From Table 3.2 of
NUREG-1275, there were 38 events that lead to a loss of pool inventory. If we do not consider
the load drop event (because this is treated separately), we have 37 events. Of these, 2 events
involved level drops greater than 5 feet. Therefore, a probability of large leak event would be
2/37 = 0.06 (6%). For the other 94% of the cases, operation of the makeup pump is sufficient to
prevent fuel uncovery.

Appendix 2 A2A-53


recorded between 1950 and 1995 (an additional 4 in the 1996 to 1998 period). It was estimated
that about 30 percent of all reported tornadoes were in the F2 to F3 range and about 2.5 percent
were in the F4 to F5 range.

The Department of Energy Report DOE-STD-1020-94, (Ref. 4) has some insights into wind-
generated missiles:

(1) For sites where tornadoes are not considered a viable threat, to account for objects or
debris a 2x4 inch timber plank weighing 15 Ibs is considered as a missile for straight winds
and hurricanes. With a recommended impact speed of 50 mph at a maximum height of
30 ft above ground, this missile would break annealed glass, perforate sheet metal siding
and wood siding up to to 3/4-in thick. For weak tornadoes, the timber missile horizontal
speed is 100 mph effective to a height of 100 ft above ground and a vertical speed of
70 mph. A second missile is considered: a 3-in diameter steel pipe weighing 75 Ibs with
an impact velocity of 50 mph, effective to a height of 75 ft above ground and a vertical
velocity of 35 mph. For the straight wind missile, an 8-in concrete masonry unit (CMU)
wall, single wythe (single layer) brick wall with stud wall, or a 4-inch concrete (reinforced) is
considered adequate to prevent penetration. For the tornado missile, an 8-to-12-in CMU
wall, single wythe brick wall with stud wall and metal ties, or a 4- to 8-inch concrete
(reinforced) slab is considered adequate to prevent penetration (depending on the missile).
(Refer to DOE-STD-1020-94 for additional details.)

(2) For sites where tornadoes are considered a viable threat, to account for objects or debris
the same 2x4 inch timber is considered but for heights above ground to 50 ft. The tornado
missiles are (1) the 15 Ibs, 2x4 inch timber with a horizontal speed of 150 mph effective up
to 200 ft above ground, and a vertical speed of 100 mph; (2) the 3-inch diameter, 75 Ibs
steel pipe with a horizontal speed of 75 mph and a vertical sped of 50 mph effective up to
100 ft above ground; and (3) a 3,000 Ibs automobile with ground speed up to 25 mph. For
the straight wind missile, an 8-in CMU wall, single wythe brick wall with stud wall, or a
4-inch concrete (reinforced) is considered adequate to prevent penetration. For the
tornado missile, an 8 in CMU reinforced wall, or a 4-to-10-inch concrete (reinforced) slab is
considered adequate to prevent penetration (depending on the missile). (Refer to
DOE-STD-1020-94 for additional details.)

3. Recommended Values for Risk-informed Assessment of SFPs
The tornado strike probabilities for each F-scale interval were determined from the SPC raw

data on a state-averaged basis. For each F-scale, the point strike probability was obtained
from the following equation:

P,, = (2«<#22] x a Equation A2e-1
Aop Yj

int
where:

P,,= strike probability for F-scale (fs)

Appendix 2E A2E-2


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3-43


APPENDIX 2E
STRUCTURAL INTEGRITY OF SPENT FUEL POOL STRUCTURES
SUBJECT TO TORNADOS

1. INTRODUCTION

Tornado damage from missiles have the potential to affect the structural integrity of the spent
fuel pool (SFP) or the availability of nearby support systems, such as power supplies, cooling
pumps, heat exchangers, and makeup water sources, and may also affect recovery actions.
Department of Energy (DOE) studies indicate that the thickness of the SFP walls (greater than
four feet of reinforced concrete) is more than sufficient protection from missiles that could be
generated by the most powerful tomadoes ever recorded in the United States. In addition, the
frequency of meeting or exceeding the wind speeds of F4 to F5 tornadoes (the most powerful
tornadoes on the Fujita scale) is estimated to be on the order of 6x10” per year in the areas of
the U.S. that are subject to the largest and most frequent tornadoes. The likelihood of meeting
or exceeding the size tornado that could damage support systems is on the order of 2x10° per
year. This is not the estimated frequency of fuel uncovery on a zirconium fire since the
frequency estimate does not include credit for maintaining pool inventory from either onsite or
offsite sources.

The probability of failing to maintain inventory was estimated for the case of loss of offsite power
from severe weather, where it was assumed that the principal impact of the severe weather was
to hamper recovery of offsite power and also to increase the probability of failing to bring offsite
sources to bear because of damage to the infrastructure. The situation with tornados is
different, because the damage caused by a tornado is relatively localized. Therefore, while a
direct hit on the plant could also disable the diesel fire pump, it would be unlikely to also disable
offsite resources to the same degree. Therefore, the probability of failing to bring in the offsite
resources can be argued to be the same as for the seismic case, i.e., 1x10, under the
assumption that NEI commitments 3 and 4 are implemented.

2. ANALYSIS

The methodology assessing tornado risk developed in NUREG/CR-2944, (Ref. 1) was used for
this evaluation. The National Climatic Data Center (NCDC) in Asheville, N.C., keeps weather
records for the U.S. for the period 1950 to 1995 (Ref. 2). Tornado data are reported as the
annual average number of (all) tornadoes per 10,000 square miles per state and the annual
average number of strong-violent (F2 to F5) tornadoes per square mile per state, as shown in
Figures A2e-1 and A2e-2.

The NCDC data were reviewed and a range of frequencies per square mile per year was
developed based on the site location and neighboring state (regional) data. In general, the
comparison of the NUREG/CR-5042 (Ref. 3) tornado frequencies for all tornadoes to the NCDC
tornado frequencies for all reported tornadoes showed good agreement between the two sets of
data.

Raw data from the Storm Prediction Center (SPC), for the period 1950 to 1995 was used to
develop a database for this assessment. About 121 F5, and 924 F4, tornadoes have been

Appendix 2E A2E-1


Table 3.1 Spent Fuel Poo! Cooling Risk Analysis — Frequency of Fuel Uncovery (per year)

Frequency of Fuel Frequency of Fuel
Uncovery (EPRI Uncovery (LLNL

INITIATING EVENT hazard) hazard)
Seismic event > 2x10°7 2x10
Cask drop © 2.0x10°7 same
Loss of offsite power’ initiated by severe 1.1x10°7 same
weather
Loss of offsite power from plant centered 2.9x10° same
and grid-related events
Internal fire 2.3x10° same
Loss of pool cooling 1.4x10% same
Loss of coolant inventory 3.0x10°° same
Aircraft impact 2.9x10 same
Tornado Missile <1.0x10° same
Total® 5.8x10° 2.4x10%

5This value is the mean of the failure probabilities for Central and Eastern SFPs that
satisfy the seismic checklist and includes seismically induced catastrophic failure of the pool
(which dominates the results) and a small contribution from seismically induced failure of pool

support systems.

°For a single-failure-proof system without a load drop analysis. The staff assumed that
facilities that chose the option in NUREG-0612 have a non-single-failure-proof system and
implemented their load drop analysis including taking mitigative actions to assure a high
confidence that the risk of catastrophic failure was less than or equivalent to that of a single-

failure-proof system.

The estimate is based upon the time available for human response when the fuel has
decayed 1 year. After only a few months of decay, these estimates are not expected to increase
significantly. Furthermore, for longer periods of decay, no significant change in the estimated
frequency is expected because the fuel handler success rates are already so high after 1 year of

decay.

®Consistent with PRA limitations and practice, contributions to risk from safeguards
events are not included in these frequency estimates. EP might also provide dose savings in

such events.

3-9


4.3.5 Top Event OFD - Operator Recovery Using Offsite Sources
4.3.5.1 Event Description and Timing

Given the failure of recovery actions using onsite sources, this event accounts for recovery of
coolant make-up using offsite sources such as procurement of a fire engine. Adequate time is
available for this action, provided that the operator recognizes that recovery of cooling using
onsite sources will not be successful, and that offsite sources are the only viable alternatives.
Fault tree LP1-OFD represents this top event for the lower branch, and LP1-OFD-U for the
upper branch. These fault trees contains those basic events from the fault trees LP1-OMK-U
and LP1-OMK-L that relate to recognition of the need to initiate the fire water system; if OMK
fails because the operator failed to recognize the need for firewater make-up, then it is assumed
that the operator will fail here for the same reason.

4.3.5.2 Relevant Assumptions

. The operators have 88 hours to provide makeup and inventory cooling.

. Procedures and training are in place that ensure that offsite resources can be brought to
bear (NEI commitments 2 and 4), and that preparation for this contingency is made when it

is realized that it may be necessary to supplement the pool make-up.

° Procedures explicitly states that if the water level drops below a certain level (e.g., 15 ft
below normal level) operator must initiate recovery using offsite sources.

* Operators have received formal training in the procedures.
¢ Offsite resources are familiar with the facility.

4.3.5.3 Quantification

Human Error Probabilities

The event HEP-INV-OFFSITE represents failure to recognize that it is necessary to take the
extreme measure of using offsite sources, given that even though there has been ample time up
to this point to attempt recovery of both the SFP cooling system and both firewater pumps it has
not been successful. This top event includes failures of both the diagnosis of the need to
provide inventory from offsite sources, and the action itself. The availability of offsite resources
is assumed not to be limiting on the assumption of an expansive preparation time. However,
rather than use a calculated HEP directly, a low level of dependence is used to account for the
possible detrimental effects of the failure to complete prior tasks successfully.

Appendix 2A A2A-40


The analyses were to include these considerations:
1. Partial draindown concerns (i.e., cases where the fuel is only partially uncovered).

2. Study of the global flow pattern in the SFP building to determine the applicability of
approximations used in previous calculations.

3. Determine the effect of detailed pool loading assumptions on critical decay times.

The staff quantified the heatup time of the fuel after uncovery as a function of the decay time
since final shutdown. The heatup time (displayed in Figure A1-1 below) was defined as the time
to heat the fuel from 30 °C to 900 °C. The heatup time of the fuel depends on the amount of
decay heat in the fuel, the oxidation heat input, and the amount of heat removal available from
the fuel. The amount of decay heat is dependent on the burnup. The amount of heat removal is
dependent on several variables that are difficult to represent generically without making a
number of assumptions that may be difficult to confirm on a plant- and event-specific basis. For
example, the air flow path is a critical parameter in the analyses. This in turn depends upon the
fuel assembly geometry, rack configuration, and loading. However, the rack configuration and
fuel assembly geometry are not only plant specific, they are subject to unpredictable changes
when subjected to a severe seismic event or cask drop accident. For this analyses the staff
initially assumed an undamaged fuel assembly and fuel rack geometry. Variations in air flow
were then used to model the effects of potential flow blockage because of damage, as plant-
specific design variations.

The staff used a specially modified version of TRAC-M to estimate the heatup time. TRAC-M
was used because it is robust, has flexible modeling capabilities, and runs fast enough to
perform sensitivity studies. Modifications were made to the wall drag, the wall heat transfer, and
the oxidation models so that they would be applicable to the SFP heatup problem. The transfer
of heat between high powered bundles and low powered bundles was not modeled, and only the
fuel and fuel rack heat structures were modeled so the heatup time estimates should be
conservative if the rack geometry is intact after the pool draining.

For the calculations, the staff used a decay heat per assembly and divided it equally among the
pins. It assumed a 9X9 assembly for the BWRs and a 17x17 assembly for the PWRs. Decay
heats were computed using an extrapolation of the decay power tables in NUREG/CR-5625
(Ref. 1). The decay heat in NUREG/CR-5625 is based on ORIGEN code calculations. The
tables used in ORIGEN for the decay heat extend to burnups of 50 GWD/MTU for PWRs and
45 GWD/MTU for BWRs. The staff recognizes that the decay heat is only valid for values up to
the maximum values in the tables, but staff ORIGEN calculations of the decay power, with
respect to burnup for values in the table, indicate that extrapolation provides a reasonable and
slightly conservative estimate of the decay heat for burnup values beyond the limits of the
tables. Current peak bundle average burnups are approximately 50 GWD/MTU for BWRs and
55 GWD/MTU for PWRs. The BWR decay heat was calculated using a specific power of

26.2 MW/MTU. The PWR decay heat was calculated using a specific power of 37.5 MW/MTU.
Both the PWR and BWR decay heats were calculated for a burnup of 60 GWD/MTU and include
an uncertainty factor of 6 percent. The pool is divided into 10 flow channels. The downcomer
flow area around the periphery of the pool is one channel. The last core offloaded is

Appendix 1A A1A-3


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A4D-6

Appendix 4D


Table A4-11 Mean Consequences for Case 5

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 96 48,300 2,260
0-500 96 449,000 20,200

90 days 0-100 83 47,500 2,220
0-500 83 460,000 20,700

1 year 0-100 67 46,700 2,180
0-500 67 473,000 21,300

As noted above, NUREG/CR-6451 estimated the release of lanthanum and cerium to be a
factor of six higher than that originally estimated in NUREG/CR-4982. Case 6 was performed
to assess the potential impact of that higher release. The Case 6 consequence results were
identical to those of Case 5 shown in Table A4-11. Therefore, even it were possible for fuel
fines to be released offsite, there would be no change in offsite consequences as a result.

The final case, Case 7 was performed to examine the impact of a 99.5% evacuation for a case
with evacuation before the release begins. This sensitivity (see Table A4-12) showed an order
of magnitude decrease in the prompt fatalities. Again, as expected, no change in the societal

dose or cancer fatalities was observed.

Table A4-12 Mean Consequences for Case 7

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 .096 48,100 2,250
0-500 .096 449,000 20,200

90 days 0-100 083 47,400 2,210
0-500 .083 460,000 20,700

1 year 0-100 .067 46,600 2,170
0-500 .067 473,000 21,300

Comparison with Earlier Consequence Analyses

As a check on the above calculations and to provide additional insight into the consequence
analysis for severe SFP accidents, the above calculations were compared to the consequence
results reported in NUREG/CR-4982 and NUREG/CR-6451. Table A4-13 shows the analysis
assumptions used for BWRs in these earlier reports together with those of Cases 3 and 4 of the

Appendix 4

A4-12



Table 5.1 Summary of Results

Initiating Event Fuel Uncovery
Frequency (per year)

Internal Fires 2.3E-08

Loss of Cooling 1.4E-08

Loss of Inventory 3.0E-09

Loss of Offsite Power (plant | 2.9E-8
centered & grid-related

events)

Loss of Offsite Power 1.1E-7
(severe weather events)

TOTAL 1.8E-07

6.0 REFERENCES

1.

8.

U.S. Nuclear Regulatory Commission, "Revised Livermore Seismic Hazard Estimates for
69 Nuclear Power Plant Sites East of Rocky Mountains," NUREG-1488, P. Sobel, October
1993.

Electric Power Research Institute, "Seismic Hazard Methodology for the Central and
Eastern United States," EPRI NP-4726, November 1988.

Memorandum, G. M. Holahan (NRC) to J. A. Zwolinski (NRC), "Preliminary Draft Technical
Study of Spent Fuel Pool Accidents for Decommissioning Plants," June 16, 1999.

Letter from L. Hendricks of the Nuclear Energy Institute (NEI) to R. Barrett of the USNRC,
November 12, 1999.

Letter, J. A. Lake (INEEL) to G. B. Kelly (NRC), "Details for the Spent Fuel Pool Operator
Dose Calculations," CCN# 00-000479, October 20, 1999.

K. D. Russell, et al., "Systems Analysis Programs for Hands-on Integrated Reliability
Evaluations (SAPHIRE), Version 5.0: Technical Reference Manual," NUREG/CR-6116,
July 1994.

Williams, J. C., "A Data-Based Method for Assessing and Reducing Human Error to
Improve Operational Performance”, in Proceedings of the 1988 IEEE Conference on
Human Factors and Poer PLants, Monterey, Ca., June 5-9, 1988, pp 436-450, Institiute of
Electrical and Electronics Engineers, New York, NY, 1988.

Holinagel, E., "Cognitive Reliability and Error Analysis Method - CREAM" Elsevier, 1998.

Appendix 2A A2A-68


Table 3.7-2 Consequences of an SFP Accident With a Low Ruthenium Source Term (per

event)
Mean Consequences for Low Ruthenium Source Term (Surry population,
95% evacuation)
Individual Risk*
Societal Dose Individual Risk” } of Latent Cancer
Time After (p-rem within 50 | of Early Fatality | Fatality (within
Shutdown Early Fatalities miles) (within imile) 10 miles)
Late Evacuation

30 days 2 5.58x10° 4.27x10? 1.88x10?
90 days 4 5.43x10° 9.86x10° 1.82x107
1 year 1 5.28x10° 7.13x10% 1,.68x10?

2 years - 5.12x106 564x107 1.58x107
5 years - 4.90x10° 3.18x10° 1.43x10?
10 years = 4.72x10° 1,63x10° 4.29x10?

Early Evacuation

30 days - 4.12x10° 8.36x10* 9.92x10
90 days + 4.02x10° 6.83x10* 9.62x10
4 year - 3.95x10° 5.44x10* 9.09x10*

2 years - 3.87x10° 4.41x10% 871x104
5 years - 3.77x10® 2.54x104 8.14x10*
10 years - 3.69x10° 1.47x10+ 7.70x10*

* Conditional on event - Total frequency for all events is shown in Table 3.1 as less than 3x10° per year.

The consequences in Table 3.7-1 are based on the upper bound source term described in
Appendix 4B. With the exception of ruthenium and fuel fines, the release fractions are from
NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants” (Ref. 1), and
include the ex-vessel and late in-vessel phase releases. The ruthenium release fraction is for a
volatile fission product in an oxidic (rather than metallic) form. This is consistent with the
experimental data reported in Reference 8. The source term is considered to be bounding for
several reasons. First, rubbling of the spent fuel after heatup to about 2500 °K is expected to
limit the potential for ruthenium release to a value less than that for volatile fission products.
Second, following the Chernobyl accident, ruthenium in the environment was found to be in the
metallic form (Ref. 2). Metallic ruthenium (Ru-106) has about a factor of 50 lower dose
conversion factor (rem per Curie inhaled) than the oxidic ruthenium assumed in the Melcor
Accident Consequence Code System (MACCS) calculations. Finally, the fuel fines release
fraction is that from the Chernobyl accident (Ref. 3). This is considered to be bounding because
the Chernobyl accident involved more extreme conditions (i.e., two explosions followed

3-30


10.

41.

U.S. Code of Federal Regulations, “Planning Basis For The Development Of State and
Local Government Radiological Emergency Response Plans In Support of Light Water
Nuclear Power Plants,” NUREG-0396, December 1978.

U.S. Nuclear Regulatory Commission, "Criteria for Preparation and Evaluation of
Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power
Reactors,” NUREG-0654/FEMA-REP-1, Revision 1, November 1980.

U.S. Code of Federal Regulations, “Requirements for Physical Protection of Licensed
Activities in Nuclear Power Reactors Against Radiological Sabotage,” Section 55, Part 73,
Title 10, “Energy.”

U.S. Code of Federal Regulations, “Financial Protection Requirements and Indemnity
Agreements,” Part 140, Title 10, “Energy.”

U.S. Nuclear Regulatory Commission, “Financial Protection Required of Licensees of Large
Nuclear Power Plants During Decommissioning,” SECY-93-127, July 13, 1993.

6-3


APPENDIX 3
ASSESSMENT OF THE POTENTIAL FOR CRITICALITY IN DECOMMISSIONING
SPENT FUEL POOL

3.1 INTRODUCTION

The staff criticality assessment includes both a more classical deterministic study and a
qualitative risk study. The conclusion in Section 3 of this report that criticality is not a risk
significant event, is based upon consideration of both of these studies. The deterministic study
was used to define the possible precursor scenarios and any mitigating actions. The risk study
considered whether the identified scenarios are credible and whether any of the identified
compensatory measures are justified given the frequency of the initiating scenario. This
appendix combines the risk study, discussed in Section 3, the consequences, and the report on
the deterministic criticality assessment into one location for easy reference.

3.2 Qualitative Risk Study

3.2.1 Criticality in Spent Fuel Pool

Because of the processes involved and lack of data, it was not possible to perform a
quantitative risk assessment for criticality in the spent fuel pool. Section 3.2.2 of this appendix,
is a deterministic study in which the staff performed an evaluation of the potential scenarios that
could lead to criticality and identified those that are credible. In this section, the staff provides
its qualitative assessment of risk because of criticality in the SFP, and its conclusions that the
potential risk from SFP criticality is sufficiently small.

In section 3.2.2, the staff evaluated the various potential scenarios that could result in
inadvertent criticality. This assessment identified two scenarios as credible, which are listed
below.

(1) A compression or buckling of the stored assemblies could result in a more optimum
geometry (closer spacing) and thus, create the potential for criticality. Compression is not
a problem for high-density PWR or BWR racks because they have sufficient fixed neutron
absorber plates to mitigate any reactivity increase, nor is it a problem for low-density PWR
racks if soluble boron is credited. But, compression of a low-density BWR rack could lead
to a criticality since BWR racks contain no soluble or solid neutron absorbing material.
High-density racks are those that rely on both fixed neutron absorbers and geometry to
control reactivity. Low-density racks rely solely upon geometry for reactivity control. In
addition, all PWR pools are borated, whereas BWR pools contain no soluble absorbing
material. If both PWR and BWR pools were adequately borated, criticality would not be
achievable for a compression event.

(2) If the stored assemblies are separated by neutron absorber plates (e.g., Boral or Boraflex),
loss of these plates could result in a potential for criticality for BWR pools. For PWR pools,
the soluble boron would be sufficient to maintain subcriticality. The absorber plates are
generally enclosed by cover plates (stainless steel or aluminum alloy). The tolerances

Appendix 3 A3-1


Table 4 - Comparison of Risk Increase with RG 1.174 Guideline (at one year)

Risk Measure

Risk Increase Because of EP Relaxation

RG 1.174 Guideline

(per year) Risk increase
— —— (per year)
Baseline ' Seismic Sensitivity?
Early Fatalities 1.5x10° 1.6x10* 2.5x10*
Population Dose 1.6 17.6 11
Individual Early 6.6x10° 7.3x10° 8.7x10*
Fatality Risk
Individual Latent 1.6x10* 1.8x107 6.9x10*
Cancer Fatality Risk

1 - Assumes no effective evacuation in seismic events, regardless of pre-planning

2  - Assumes maximum effectiveness of emergency planning (i.e., early evacuation) when EP
requirements are maintained, and minimum effectiveness (i.e., late evacuation) when EP
requirements are relaxed

Appendix 4D

A4D-7



current analysis.

NUREG/CR-4982 results included consequence estimates for societal dose for an operating
reactor for severe SFP accidents occurring 30 days and 90 days after the last discharge of
spent fuel into the pool. The Case 3 results were compared against the NUREG/CR-4982
results, because they use the same population density (100 persons/mile?) and 11 batches of
spent fuel in the pool. However, one difference is that Case 3 uses a radionuclide inventory
that is a factor of 1.7 higher than NUREG/CR-4982 to reflect the relative power levels of a large
BWR and Millstone 1. Therefore, Case 3 was rerun with the radionuclide inventory of

NUREG/CR-4982. As shown in Table A4-14, the Case 3 rerun results generally compared well
with the NUREG/CR-4982 results.

Appendix 4 A4-13


FALLRE TOREOOER
(COOUNSSTSTEN

CPERATOREALS TO
FEST

Q.

FALURETOREONER
‘COOUNG SYSTEM

PERATORFALS TO
IE SPOLAT
NO DEPENDENCE,

A3



Ref. 5:

Vermont Yankee (BWR): Cucrpr = 0.48g PGA

Q)
Robinson (PWR): Cucupr = 0.65g PGA

These two fragility estimates provide some verification of the HCLPF capacity assumption of
0.4 to 0.5g PGA used in Ref. 1.

lam confident that a set of seismic screening tables (seismic check lists) can be
developed to be used with seismic walkdowns and drawing reviews to provide reasonable
assurance that the HCLPF capacity of spent fuel pools is at least in the range of 0.4 to 0.5g PGA
for spent fuel pools that pass such a review. However, in order to justify a HCLPF capacity in
the range of 0.4 to 0.5g PGA, these screening tables will have rather stringent criteria so that I
am not so confident that the vast majority of spent fuel pools will pass the screening criteria.
The screening criteria (seismic check lists) summarized in Ref. 4 provides an excellent start. The
subject of screening criteria is discussed more thoroughly in Section 3.

Once the HCLPF seismic capacity (Cycrpr) has been estimated, the seismic risk of failure
of the spent fuel pool can be estimated by either rigorous convolution of the seismic fragility
(conditional probability of failure as a function of ground motion level) and the seismic hazard
(annual frequency of exceedance of various ground motion levels), or by a simplified
approximate method. This subject is discussed more thoroughly in Ref. 10.

A simplified approximate method is used in Ref. 1 to estimate the annual seismic risk of
failure (P;) of the spent fuel pool given its HCLPF capacity (Cucipr). The approach used in Ref.
1 is that:

P; = 0.05 Hucure @G)

where Hycipr is the annual frequency of exceedance of the HCLPF capacity. Ref. 1 goes on to
state that for most Central and Eastern U.S. (CEUS) plants, the mean annual frequency of
exceeding 0.4 to 0.5g PGA is on the order of or less than 2x10° based on the Ref. 8 hazard
curves. Thus, from Eqn. (3), the annual frequency of seismic-induced gross failure (P;) of the
spent fuel pool is on the order of 1x10° or less for most CEUS plants.

Unfortunately, the approximation of Eqn. (3) is unconservative for CEUS hazard curves
that have shallow slopes. By shallow slopes, I mean that it requires more than a factor of 2
increase in ground motion to correspond to a 10-fold reduction in the annual frequency of
exceedance. For most CEUS sites, Ref. 8 indicates that a factor of 2 to 3 increase in ground
motion is required to reduce the hazard exceedance frequency from 1x10* to 1x10*. Over this
range of hazard curve slopes, Eqn. (3) is always unconservative and will be unconservative by a
factor of 2 to 4. Therefore, a HCLPF capacity in the range of 0.4 to 0.5g PGA is not sufficiently
high to achieve a spent fuel pool seismic risk of failure on the order of 1x10° or less for most

A2B-12


Response: The staff agrees with the commenter that the study did not address the topics of
relocation and societal impacts, such as land interdiction. As part of its original licensing
review, every operating plant had an environmental impact statement that addressed land use
for the area surrounding that plant. When a plant enters decommissioning, an environmental
assessment is performed to determine whether activities will remain bounded by that
environmental impact statement.

The calculations in support of this risk study were performed following the principles and
approach of Regulatory Guide (RG) 1.174, "An Approach For Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"
which does not include environmental considerations. While overall societal risk is not
considered directly in RG 1.174, a large early release is used to gauge the severity of the event
outside of the plant boundary. Similarly, in this study, early and latent effects were directly
calculated and reported.

The Commission recently considered whether an additional agency safety goal or objective
was needed to directly address land contamination and overall societal risk. It was decided by
the Commission that the current policy would not change. For further discussion, read
SECY-00-0077, dated March 30, 2000, and staff requirements memorandum dated

June 27, 2000. Consistent with Commission guidance, the staff does not plan to include this
issue in the study.

Appendix 6 A6-36


To evaluate the risk from a seismic event at an SFP, one needs to know both the likelihood of
seismic ground motion at various g-levels (i.e., seismic hazard) and the conditional probability
that a structure, system, or component (SSC) will fail at a given acceleration level (i.e., the
fragility of the SSC). These are convolved mathematically to arrive at the likelihood that the SFP
will fail from a seismic event. In evaluating the effect of seismic events on SFPs, it became
apparent that although information was available on the seismic hazard for nuclear power plant
sites, the staff did not have fragility analyses of the pools, nor generally did licensees. The staff
recognized that many of the SFPs and the buildings housing them were designed by different
architect engineers. The SFPs and structures housing them were built to codes that evolved
through various code editions.

The staff originally performed a simplified bounding seismic risk analysis in its June 1999 draft
assessment of decommissioning plant risks to help determine if there might be a seismic
concern. The analysis indicated that seismic events could not be dismissed on the basis of a
simplified bounding approach. In addition after further evaluation and discussions with
stakeholders, it was determined that it would not be cost effective to perform a detailed plant-
specific seismic evaluation for each SFP. Working with its stakeholders, the staff developed
other tools that help assure the pools are sufficiently robust.

2. RETURN PERIOD OF SFP-FAILING EARTHQUAKE GROUND MOTIONS

The staff reexamined its methods for estimating seismic risk and reexamined the results of
Table 3 in Attachment 2 to this appendix, which estimates the return frequencies of large
earthquake ground motions that could fail SFPs. It was decided that the HCLPF value of 1.2 g
peak spectral acceleration was a good generic screening measure for seismic capacity for
decommissioning plant SFPs. The staff used a simplified, but slightly conservative method (see
Attachment 2) to estimate the annual probability of a zirconium fire because of seismic events
(including use of site-specific seismic hazard estimates). These calculations resulted in a range
of site-specific frequencies from less than 1x10* per year to over 1x10° per year, depending on
the site and the seismic estimates used.

Both the EPRI and LLNL hazard estimates are judged by the NRC to be equally valid and useful
for making decisions. At most sites, the LLNL hazard estimates predict a higher frequency than
EPRI estimates for ground motions exceeding a given acceleration. From a regulatory stand
point, it is prudent to examine the implications of the more limiting parameter inputs (e.g.,
shorter return period of large earthquakes) when making safety decisions. Using the LLNL
hazard estimates, all central and eastern sites except H. B. Robinson have an expected
seismic-induced zirconium cladding fire frequency less than the Pool Performance Guideline
(1x10° per year) at a 1.2g peak spectral acceleration (PSA) ground motion. Using EPRI hazard
estimates, only one site east of the Rocky Mountains would have an expected frequency of a
seismic-induced cladding fire in excess of 1x10* per year(but less than 1x10° per year) at the
1.2g PSA level. Neither LLNL nor EPRI developed hazard estimates for sites west of the Rocky
Mountains (such as San Onofre, Diablo Canyon, and WNP2), and EPRI, unlike LLNL, did not
develop hazard estimates for all operating reactor sites east of the Rockies. H.B. Robinson and
Western sites with higher frequencies at the 1.2g PSA level would need to perform site-specific
seismic risk analyses for their SFPs if the utilities desire to take advantage of potential
exemptions or rule changes regarding EP, security, or indemnification. The plant-specific

Appendix 2B A2B-2


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Cs-137 30.0y 2.02E7 2.01E7 1.97E7
Ba-140 12.8d 5.19E6 1.90E5 6.41E-2
La-140 12.8d 5.97E6 2.19E5 7.37E-2
Ce-141 32.5d 1.32E7 3.61E6 1.03E4
Ce-144 284.6d 2.64E7 2.27E7 1.16E7
Pr-143 13.6d 5.44E6 2.41E5 1.90E-1
Nd-147 11.0d 1.5466 3.36E4 1.10E-3
Np-239 2.4d 5.59E4 2.88E3 2.88E3
Pu-238 87.7y 4,51E5 4.53E5 4.54E5
Pu-239 24100y 8.89E4 8.89E4 8.89E4
Pu-240 6560y 1.30E5 1.30E5 1.30E5
Pu-241 14.4y 2.29E7 2.27E7 2.19E7
Am-241 432.7y 2.88E5 2.94E5 3.21E5
Cm-242 162.8d 1.45E6 1.12E6 3.50E5
Cm-244 18.1y 2.27E5 2.25E5 2.19E5

MACCS has a default list of 60 radionuclides that are important for offsite consequences for
reactor accidents. NUREG/CR-4982 contains inventories for 40 of these 60 radionuclides. Of
these 40 radionuclides, 27 have half-lives from 2.4 days to a year and 13 have half-lives of a
year or greater as shown in Table A4-1. The half-lives of the remaining 20 radionuclides range
from 53 minutes to 1.5 days as shown in Table A4-2. Because the largest half-life of these 20
radionuclides is 1.5 days, omitting these 20 radionuclides from the initial inventories used in the
MACCS analysis should not affect doses from releases occurring after a number of days of

decay.

Appendix 4

A4-4



7.0 ACRONYMS

ACRS
ANSI
ANS
ASB
atm

BNL
BTP
BWR

CFD
CFM
CFR

DOE
DSP

ECCS
EOF
EP
EPRI
EPZ
ET

FFU
FT

gpm
Gl

GWD

HCLPF
HRA
HVAC

Ipc
INEEL
ISFSI

kw
LERF

LLNL
LOSP

Advisory Committee on Reactor Safeguards

American National Standard Institute

American Nuclear Society

NRC Auxiliary Systems Branch (Plant Systems Branch)
atmosphere

Brookhaven National Laboratory
branch technical position
boiling-water reactor

computational fluid dynamics
cubic feet per minute
U.S. Code of Federal Regulations

Department of Energy
decommissioning status plant

emergency core cooling system
emergency operations facility
emergency plan

Electric Power Research Institute
emergency planning zone

event tree

frequency of fuel uncovery
fault tree

gallon(s) per minute
generic issue
gigawatt-day

high confidence in low probability of failure
human reliability analysis
heating, ventilation, and air conditioning

industry decommissioning commitment
Idaho National Engineering and Environmental Laboratory
independent spent fuel pool installation

kilowatt

large early release frequency

Lawrence Livermore National Laboratory
loss of offsite power

light-water reactor

7-1


August 25, 2000

MEMORANDUM TO: Gary M. Holahan, Director
Division of Systems Safety and Analysis
Office of Nuclear Reactor Regulation

FROM: Farouk Eltawila, Acting Director
Division of Systems Analysis and Regulatory Effectiveness
Office of Nuclear Regulatory Research

SUBJECT: RISK-INFORMED REQUIREMENTS FOR DECOMMISSIONING

As part of its effort to develop generic, risk-informed requirements for decommissioning, NRR
requested (Reference 1) an evaluation of the offsite radiological consequences of beyond-
design-basis spent fuel pool accidents. In response to that user need, we completed an in-
house analysis (Reference 2) that concluded the following:

. The short-term consequences (i.e., early fatalities) decreased by a factor of two when
the fission product inventory decreased from that for 30 days to that for one year after
final shutdown.

. At one year after final shutdown, the short-term consequences decreased by up to a
factor of 100 as a result of early evacuation. Early evacuation is likely after one year,
because of the decreased decay heat level and the number of hours required for the
fuel with the highest decay power to heat up to the point of releasing fission products.

. The long-term consequences (i.e., cancer fatalities and societal dose) were unaffected
by the additional decay and early evacuation.

Although the reductions in the short-term consequences were significant, emergency planning
requirements could not be relaxed solely on the basis of these reductions. NRR also used our
consequence evaluation in the Draft Final Technical Study of Spent Fuel Pool Accident Risk at
Decommissioning Nuclear Power Plants, February 2000, as an absolute measure of spent fuel

pool accident consequences and concluded that the consequences were generally comparable
to those of reactor accidents.

Subsequently, the ACRS raised issues with the source term and plume modeling associated
with spent fuel pool accidents. In particular, the ACRS believed that the ruthenium and fuel
fines releases and plume spreading were too low. To address these issues, we completed a
series of sensitivity studies and concluded:

° With the exception of the ruthenium release fraction, the parameters varied did not
sufficiently impact the results, nor change the conclusion that the consequences were
generally comparable to those of reactor accidents.

° Increasing the ruthenium release fraction from that for a non-volatile (2x10) to that for
a volatile (.75) resulted in a large increase in both short-term and long-term
consequences due to ruthenium’s high dose per curie inhaled. However, consequence
increases from ruthenium were demonstrated to be largely offset by early evacuation.

: Although using updated values for plume-spreading model parameters resulted in up to
a 60% increase in long-term consequences, similar increases are expected when these
updated values are used to calculate reactor accident consequences. Using updated
values also resulted in up to a factor-of-15 decrease in short-term consequences.


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Appendix 4D


Comment #96: Although NRC and EPA disagree on site remediation criteria, the commenter
stated that either level. would provide reasonable assurance to the public of undue risk.

Response: Resolution of the disagreement between NRC and EPA on release criteria is not
within the scope of the technical study.

Comment #97: What is the applicability of 10 CFR Part 26 fitness-for-duty regulations to
decommissioning reactors?

Response: Fitness-for-duty at decommissioning facilities is one of the issues that will be
evaluated by the decommissioning regulatory improvement initiative.

Comment #98: Quality assurance, emergency planning, fire protection, and application of
codes and standards differs from site to site. Right now the decommissioning industry is being
regulated by exemption to Part 50.

Response: The staff is planning to propose new emergency planning rules for
decommissioning reactors to eliminate the need for addressing the issue on a plant-specific
basis by processing exemptions. A final regulatory guide on decommissioning reactor fire
protection programs is expected to be issued in 2001. The remaining issues will be addressed
by the decommissioning regulatory improvement initiative. The planned rulemaking and
guidance should assist in making regulations predictable and consistent.

Comment #99: The issue of on-site disposal of clean waste (rubblization) needs clarification.

Response: The development of NRC policy on rubblization is now ongoing in the Office of
Nuclear Materials Safety and Safeguards.

Comment #100: A commenter requested an adjudicatory hearing and a prior NRC
review/approval step at the onset of the decommissioning process.

Response: This issue of a hearing and NRC review and approval prior to decommissioning
has been previously considered by the Commission. The Commission addressed the issue in
the statements of consideration for the rulemaking for decommissioning published July 29,
1996, in the Federaf Register (61 FR39278) by stating: “...initial decommissioning activities
(dismantlement) are not significantly different from routine operational activities such as
replacement or refurbishment. Because of the framework of regulatory provisions embodied in
the licensing basis for the facility, these activities do not present significant safety issues for
which an NRC decision would be warranted.” Therefore, an NRC review and approval process
that allows a public hearing before decommissioning begins is not necessary. However, in the
1996 rulemaking the Commission decided to offer a public hearing opportunity later in the
decommissioning process at the license termination stage when issues such as to the
adequacy of site cleanup could be raised.

Comment #101: A commenter felt that the NRC should hire a contractor to determine why/how
10 CFR Part 50 was contorted to fit decommissioning reactors with the duct tape of
10 CFR 50.82 to avoid adjudicatory processes with regulatory handles.

Appendix 6 A6G-32


4.1.3.3 Quantification
Human Error Probabilities

The functional fault trees include two human failure events, depending on whether the control
room alarms have failed, or whether there was a failure to respond to the initial alarm ( it is
assumed that the alarm was canceled). If the operator failed to respond to control room alarms,
then event HEP-WLKDWN-DEPEN models subsequent operating crews’ failures to recognize
the loss of cooling during walk-downs, taking into account the dependence on event
HEP-DIAG-ALARM. A specific mechanism for dependence can only be identified on a plant and
event specific basis, but could result, for example, from an organizational failure that leads to
poor adherence to plant procedures. Because this is considered unlikely, and because the
conditions in the pool area change significantly over the time scale defined by the success
criterion for this event, the degree of dependence is assumed to be low.

If the alarms failed, then event HEP-WLKDWN-LSFPC models subsequent crews' failures to
recognize the loss of cooling during walk-downs, with no dependence on previous HEPs.
However, because the control room readouts could share a dependency with the alarms, the
assumption of local temperature measurements becomes important. The failure probabilities for
these events were developed using THERP, and are based upon three individual failures: failure
to carry out an inspection, missing a step in a written procedure, and misreading a measuring
device. Because there are on the order of 33 - 43 hours before the SFP cooling system
becomes irrecoverable without pool make-up, it is assumed that multiple crews would have to
fail. Assuming that the crews are totally independent would give a very low probability.
However, a low level of dependence is assumed and the probability is truncated at 1E-05.

4.1.3.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-WLKDWN-LSFPC. 1.0E-5
HEP-WLKDWN-DEPEN 5.0E-2

4.1.4 Top Event OCS — Operator Recovery of Cooling System
4.1.4.1 Event Description and Timing

Once the operators recognize loss of SFP cooling, they will likely focus their attention on
recovery of the SFP cooling system. It is assumed that only after bulk boiling begins and the
water level drops below the cooling system suction that the operator will inject water from other
makeup systems (e.g., firewater). Therefore, the time available to recover the SFP cooling
system could be as long as 43 hours, given an immediate response to an alarm. However, it
has been assumed that the operating staff has only until shortly after bulk boiling begins
(assumed to be 33 hours) to restore the SFP cooling system. This assumption is based on
concerns about volume reduction because of cooling and whether the makeup system capacity
is sufficient to overcome that volume reduction.

The initial cause of the loss of cooling could be the failure of a running pump in either the

Appendix 2A A2A-17


ATTACHMENT B

SPAR HRA Worksheet


4.1.5.2 Relevant Assumptions

. The operators have 88 hours to provide make-up.

. The operators will avoid using raw water (e.g., water not chemically controlled) if possible.

. The boil-off rate is assumed to be higher than the SFP makeup system capacity.

. The operators are aware that they must use raw water to refill the pool once the level
drops to below the suction of the cooling system and the pool begins boiling, since the

makeup system cannot compensate for the boiling.

. For repair of failed pumps, it is assumed that it takes 16 hours to contact maintenance
personnel, identify the problem, and get new parts.

. There is a means to remotely align a makeup source to the SFP without entry to the refuel
floor, so that make-up can be provided even when the environment is uninhabitable
because of steam and/or high radiation (NEI commitment no.8).

. Repair crew is different than onsite operators.

. Mean time to repair the firewater pump is 10 hours.

° Operators have received formal training and there are procedures that include clear
guidance on the use of the firewater system as a makeup system (NEI commitment no. 2).

. Firewater pumps are maintained and tested on a regular schedule (NEI commitment
no. 10).

4.1.5.3 Quantification

Human Error Probabilities

Three human failure events are modeled in functional fault tree LOC-OFD
HEP-RECG-FWSTART represents the operator's failure to recognize the need to initiate the
firewater system. The conditions under which the firewater system is to be used are assumed
to be explicit in a written procedure. This event was quantified using the SPAR HRA technique.
The assumptions include expansive time (> 24 hours), a high level of stress, diagnostic type
procedures, good ergonomic interface, and good quality of work process. This diagnosis task
provides the diagnosis for the subsequent actions taken to re-establish cooling to the pool.

HEP-FW-START represents failure to start the electric or diesel firewater pump within 88 hours
after the onset of bulk boiling, given that the decision to start a firewater pump was made. No
difficult valve alignment is required. This event was quantified using SPAR HRA technique. An
expansive time (> 50 times the required time), high stress, highly complex task because of its
non-routine nature, quality procedures available, as well as good ergonomics including

Appendix 2A A2A-20


For a non-single-failure-proof load handling system, the drop frequency of a heavy load drop is
estimated, based on NUREG-0612 information, to have a mean value of 3.4x10% per year. The
number of heavy load lifts was based on the NEI estimate of 100 spent fuel shipping cask lifts
per year, which probably is an overestimate. For plants with a single-failure-proof load handling
system or plants conforming to the NUREG-0612 guidelines, the drop frequency is estimated to
have a mean value of 9.6x10* per year, again for 100 heavy load lifts per year but using data
from U.S. Navy crane experience. Once the load is dropped, the analysis must then consider
whether the drop significantly damages the SFP.

When estimating the failure frequency of the pool floor and pool wall, the staff assumes that
heavy loads travel near or over the pool approximately 13 percent of the total path lift length (the
path lift length is the distance from where the load is lifted to where it is placed on the pool floor).
The staff also assumes that the critical path (the fraction of total path the load is lifted high
enough above the pool to damage the structure in a drop) is approximately 16 percent. The
staff estimates the catastrophic failure rate from heavy load drops to have a mean value of
2.1x10° per year for a non-single-failure-proof system relying on electrical interlocks, fuel
handling system reliability, and safe load path procedures. The staff estimates the catastrophic
failure rate from heavy load drops to have a mean value of 2x10” per year for a single-failure-
proof system. The staff assumes that licensees that chose the non-single-failure-proof system
option in NUREG-0612 performed appropriate analyses and took mitigative actions to reduce
the expected frequency of catastrophic damage to the same range as for facilities with a single-
failure-proof system.

NEI has made a commitment (IDC #1) for the nuclear industry that future decommissioning
plants will comply with Phases | and II of the NUREG-0612 guidelines. Consistent with this
industry commitment, the additional assurance of a well-performed and implemented load drop
analysis, including mitigative actions, is assumed in order to arrive at an accident frequency for
non-single-failure-proof systems that is comparable to the frequency for single-failure-proof
systems.

SDA#5_ Load Drop consequence analyses will be performed for facilities with non-single-
failure-proof systems. The analyses and any mitigative actions necessary to preclude
catastrophic damage to the SFP that would lead to a rapid pool draining should be
performed with sufficient rigor to demonstrate that there is high confidence in the
facility's ability to withstand a heavy load drop.

Although this study focuses on the risk associated with wet storage of spent fuel during
decommissioning, the staff has been alert to any implications for the storage of spent fuel during
power operation. With regard to power operation, the resolution of Generic Issue (Gl) 82,
“Beyond Design Basis Accidents in Spent Fuel Pools,” and other studies of operating reactor
SFPs concluded that existing requirements for operating reactor SFPs are sufficient. In
developing the risk assessment for decommissioning plants, the staff evaluated the additional
issue of a drop of a cask on the SFP floor rather than just on a SFP wall. As noted above,
because the industry has committed to Phase Il of NUREG-0612, “Control of Heavy Loads at
Nuclear Power Plants, Resolution of Generic Technical Activity A-36," this is not a concern for
decommissioning reactors.

3-17


Comment #22: The initiating-frequencies, human error rates, and equipment failure rates
should more accurately take into account the occurrence of actual events such as Chernobyl
and Three Mile Island.

Response: The decommissioning SFP risk assessment takes into account actual events that
are applicable to spent fuel pools and their support systems. The staff used initiating event
frequencies from staff studies from actual events at spent fuel pools, actual crane lift data,
site-specific seismic hazard curves, studies on aircraft crashes and tornadoes, and large
databases developed to provide estimates for initiating events and equipment failure rates.
Human error rates were developed by the staff in conjunction with experts at Idaho National
Engineering and Environmental Laboratory. The staff believes that the values used in the
report provide a reasonable picture of the risks associated with operation of decommissioning
spent fuel pools under the assumptions and commitments documented in the study.

Comment #23: The NRC should determine which failure rates used in the report are reliable
and which are not, and the results should be included in the study.

Response: The staff uses the most reliable information on failure rates that is available.
Because of the long time it takes for water above the spent fuel to heat up and boil off, the
failure rates of specific equipment that support a spent fuel pool are not important contributors
to spent fuel pool risk for long term sequences (i.e., the results are not particularly sensitive to
the assumed failure rate of equipment.) However, very large seismic events or heavy load
drops could rapidly drain the spent fuel pool. For seismic events, the robustness of the spent
fuel pool is assured by implementation of a seismic checklist (See Appendix 2b). For heavy
load drops, IDC #1 and SDA #5 calls for performance of cask drop analyses/load drop
consequence analysis or use of a single-failure-proof crane when moving heavy loads over or
near the spent fuel pool, which should help assure that the risk from heavy load drops is
extremely low.

Comment #24: Mitigating systems at decommissioning spent fuel pools are not automatic. The
NRC should assure that fuel handlers are available in the event of an accident.

Response: The decommissioning rulemaking plan will address operator staffing requirements
and safeguards staffing for facilities undergoing decommissioning. Staffing at present day
decommissioning sites is controlled by Technical Specifications on a plant-specific basis. In
addition, SDA #2 calls for walkdowns of the spent fuel pool area by fuel handlers every shift.

Comment #25: What measures have been taken to assure that fuel handlers remain attentive?

Response: The Commission, through the “Policy on Factors Causing Fatigue of Operating
Personnel at Nuclear Reactors” provides guidelines on working hours that were consistent with
the objective of ensuring that the mental alertness and decision-making abilities of plant staff
were not significantly degraded by fatigue. For this study, the staff incorporated several
measures into the risk assessment to help assure fuel handler attentiveness. First, SDA #2
calls for walkdowns of the spent fuel pool area by fuel handlers every shift. Second, IDC #5
states that SFP instrumentation will be in place providing readouts and alarms in the control
room or where the fuel handlers are stationed. Additionally, discussions with the industry
indicate that it is a general practice for sites to log instrument readings from the
decommissioning spent fuel pools at least once per shift. Such practices help maintain fuel
handler alertness and keep them abreast of the status of the pool and its support systems.

Appendix 6 AG-8


Basic Event Name Description Basic Event Probability
SPC-PMP-FTF-2 SFP cooling pump 2 fails to start and 3.9E-3

run
FP-2PUMPS-FTF Failure of firewater pump system 6.7E-4
FP-DGPUMP-FTF Failure of the diesel-driven firewater 1.8E-1

pump

4.5 Loss of Inventory Event Tree

This event tree (Figure 4.5) models general loss of inventory events, that are not the result of
catastrophic failures that could result from events such as dropped loads, tornado missiles, or
seismic events. The following assumption was made in the development of the event tree.

. Maximum depth of siphon path is assumed to be 15 ft. below the normal poo! water level
(related to NEI commitments 6 and 7). Once the water level drops 15 ft below the normai
pool water level, the losses would be only from the boil-off. This assumption may be
significant, and potentially non-conservative for sites that do not adopt NEI commitments 6
and 7.

4.5.1 Initiating Event LOI - Loss of Inventory
4.5.1.1 Event Description and Timing

This initiator (IE-LO1) includes loss of coolant inventory from events such as those resulting
from configuration control errors, siphoning, piping failures, and gate and seal failures.
Operational data provided in NUREG-1275 (Ref. 14), show that the frequency of loss of
inventory events in which the level decreased more than one foot can be estimated to be less
than one event per 100 reactor years. Most of these events were the result of operator error
and were recoverable. NUREG-1275 shows that, except for one event that lasted for 72 hours,
there were no events that lasted more than 24 hours. Eight events resulted in a level decrease
of between one and five feet and another two events resulted in an inventory loss of between
five and 10 feet.

4.5.1.2 Relevant Assumption

. NEI commitments 6 and 7 will reduce the likelihood of a significant initiating event.

4.5.1.3 Quantification

The data reviewed during the development of NUREG-1275 (Ref. 14) indicated fewer than one
event per 100 years in which level decreased over one foot. This would give a frequency of 1E-
02. However, it is assumed that the NEI commitments 6 and 7 when implemented will reduce

this frequency by an order of magnitude or more. Thus the frequency is estimated as 1E-03 per
year.

Appendix 2A A2A-52


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xii


NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION $1. REPORT NUMBER
2-89)

Neo 1102, (Assigned by NRC, Add Vol., Supp., Rov.,
3201, 3202 BIBLIOGRAPHIC DATA SHEET and Addendum Numbers, if any.)

(See instructions on the reverse)
2. TITLE AND SUBTITLE NUREG-1738

Technical Study of Spent Fue! Pool Accident Risk at Decommissioning Nuclear Power Plants

3. DATE REPORT PUBLISHED

‘MONTH ] YEAR
February 2001

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

T.E. Collins, G. Hubbard Technical

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; i contractor,
provide name and mailing address.)
Division of Systems Safety and Analysis
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type “Same as above’; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,
and mailing address.)

Same as 8. Above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

This study contains the results of the NRC staff's evaluation of the potential accident risk in a spent fuel pool at
decommissioning plants in the United States. This study was prepared to provide a technical basis for decommissioning
tulemaking for permanently shutdown nuclear power plants. This study describes a modeling approach of a typical
decommissioning plant with design assumptions and industry commitments; the therma!-hydraulic analyses performed to
evaluate the behavior of spent fuel stored in the spent fuel pool at decommissioning plants; the risk assessment of spent fuel
Pool accidents; the consequence calculations; and the sensitivity study and implications for decommissioning regulatory
requirements. Preliminary drafts of this study were issued for public comments and technical reviews in June 1999 and
February 2000. Comments from interested stakeholders, the Advisory Committee on Reactor Safeguards, and other technical
teviewers have been taken into account in preparing this study. A broad quality review was also carried out at the Idaho
National Engineering and Environment Laboratory, and a panel of human reliability analysis experts evaluated the report's
assumptions, methods, and modeling. Public comments on draft versions of this study are discussed in Appendix 6 of this,
NUREG.

12, KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report) ‘13, AVAILABILITY STATEMENT

‘Spent fuel pool accident at decommssioning nuclear power plants. unlimited

(This Page)
unclassified

(This Report)
unclassified

14. SECURITY CLASSIFICATION

15. NUMBER OF PAGES

16. PRICE

J

heevune
NRC FORM 335 (2-89) This form was electronically produced by Elite Federal Forms, Inc.


Attachment 1

Seismic Check list for Commercial Nuclear Power Plants
During Decommissioning

A2B-6


Figure 4.3 Plant centered and grid related loss of offsite power event tree

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4.3.2.3 Quantification

The basic event that represents recovery of offsite power for plant-centered and grid-related
LOSP is REC-OSP-PC. The data in NUREG/CR-5496 indicates that one event in 102 plant
centered events resulted in a loss for greater than 24 hours, and all 6 of the grid centered
events were recovered in a relatively short time. The majority of the plant-centered events were
recovered within 7 hours, so even if there is a delay in bringing repair personnel onsite, there is
a high probability of recovering offsite power within 24 hours. Therefore a non-recovery
probability of 1E-02 is assumed.

4.3.2.4 Basic Event Probability

Appendix 2A A2A-35


41 1 1 2x10? 2x10* 2x10* 2x10% 1x10® 1x10®

A subsequent review of the staffs spent fuel pool risk assessment indicated that significant air
ingression, influencing fission product release, will occur in accidents involving quick drain-down, and the
staffs consequence assessment should accommodate any reasonable uncertainty in the progression of
the accident with the possible exception of an increase in the ruthenium release. The ruthenium release
fraction used in the staffs consequence assessment was 2x10°. Small-scale Canadian experiments
show that, in an air environment, significant ruthenium releases begin after the oxidation of 75% to 100%
of the cladding, and that the ruthenium release fraction can be as high as the release fraction of the
volatile fission products. However, in a spent fuel pool accident, rubbling of the fuel may limit the
ruthenium release fraction to a smaller value than that of the volatile fission products.

With regard to the number of fuel assemblies releasing fission products, the thermal-hydraulic evaluation
in the BNL study indicated that, as a result of radioactive decay, assemblies other than those from the
final core may not reach temperatures high enough to release fission products. The number of
assemblies assumed to release fission products in the Appendix 4 consequence assessment is
equivalent to 3.5 cores. With regard to the release fractions of the low-volatile and non-volatile fission
products, higher release fractions than those in the BNL study may be possible as a result of the release
of fuel fines due to fuel pellet decrepitation associated with high fuel burnup.

Ruthenium:

To assess the sensitivity of the consequences to the ruthenium release fraction, the staff performed
consequence calculations with and without significant ruthenium releases. The starting point for this
assessment was the Base Case calculation from Appendix 4. Then, sensitivity cases were run with a
ruthenium release fraction of one and a uniform population density of 100 people/mile*. The results of
these cases (j.e., Base Case, Cases 11, 21, 22) are given in Table 3. For these cases, the effect of
ruthenium is to increase the number of prompt fatalities by a factor of ten to 90. The effect on societal
dose and cancer fatalities is a more modest increase, with the largest effect being a factor-of-four
increase in cancer fatalities for the Surry population density.

Table 3 Results of Ruthenium Release Sensitivities

(99.5% evacuation)
Case Population Ruthenium Mean Consequences (within 100 miles)
Density? release
fraction Prompt Societal Dose Cancer Fatalities
Fatalities (person-rem)

Base Case | Surry 2x10° 4.01 4.54x10° 2,320
1 Surry 1 95.3 9.53x10° 9,150
21 uniform 2x10 9.33 5.05x10® 2,490
22 uniform 1 134 9.46x10® 6,490
13° Surry 2x10* 0048 4.18x10® 1,990
14° Surry 1 132 6.75x10® 6,300
15? uniform 2x10° 045 4.65x10° 2,170
16* uniform 1 277 6.38x10° 4,940

*Based on evacuation before release.
>The uniform population density site has a population density of 100 people/mile? with an Exclusion Area
Boundary of .75 miles.

-2-


Basic Event Name Description Probability
SPC-PMP-CCF SFP cooling pumps - CCF 5.9E-4
SPC-PMP-FTF-1 SFP cooling pump 1 fails to start and run 3.9E-3
SPC-PMP-FTF-2 SFP cooling pump 2 fails to start and run 3.9E-3

4.4 Severe Weather Loss of Offsite Power Event Tree

This event tree represents the loss of SFP cooling resulting from a loss of offsite power from
severe-weather-related events. Until offsite power is recovered, the electrical pumps would be
unavailable, and only the diesel fire pump would be available to provide make-up.

Figure 4.4 shows the Severe Weather Loss of Offsite Power (LOSP) event tree sequence
progression.

4.4.1 Initiating Event LP2 - Severe Weather Loss of Offsite Power
4.4.1.1 Event Description

Initiating event IE-LP2 represents severe-weather-related losses of offsite power. Severe
weather threatens the safe operation of an SFP facility by simultaneously causing loss of offsite
power and potentially draining regional resources or limiting their access to the facility. This
event tree also differs from the plant-centered and grid-related LOSP event tree in that the
probability of offsite power recovery is reduced.

4.4.1.2 Quantification

The LOSP frequency from severe weather events is 1.1E-2/yr, taken from NUREG/CR-5496
(Ref. 18). This includes contributions from hurricanes, snow and wind, ice, wind and salt, wind,
and one tornado event, all of which occurred at a relatively small number of plants. Therefore,
for the majority of sites, this frequency is conservative, whereas, for a few sites it is non-
conservative. Because of their potential for severe localized damage, tornados were analyzed
separately in Appendix 2E.

4.4.2 Top Event OPR — Offsite Power Recovery

4.4.2.1 Event Description and Timing

The fault tree for this top event (LP2-OPR) is a single basic event that represents the
non-recovery probability of offsite power. It is assumed that if power is recovered before boil-off

starts (33 hours), the operator has a chance to reestablish cooling using the SFP cooling
system.

Appendix 2A A2A-43


<a>,= tornado area, mi?

Ag = area of observation, mi? (state land area)

Yint = interval over which observations were made, years
=n = sum of reported tornados in the area of observation

The tornado area, <a>;, was evaluated at the midpoint of the path-length and path-width
intervals shown in Table A2e-1, based on the SPC path classifications. For example, an F2
tornado with a path-length scale of 2 has an average path length of 6.55 miles and with a path-
width scale of 3, an average width of 0.2 miles.

The tornado area, <a>,, was then modified using the method described in NUREG/CR-2944
(based on Table 6b and 7b) to correct the area calculation by observations of the variations in a
tornado’s intensity along its path length and path width (see Figure A2e-3). Table A2e-2 gives
the path-length correction data. Table A2e-3 gives the path-width correction data. The
corrected effective area has a calculated <a>, of about 0.28 mi?. The combined variation in
intensity along the length and across the width of the tornado path is shown in Table A2e-4
(Table 15b from NUREG/CR-2944). For example, an F2 tornado with a path-length scale of 2
and a path-width scale of 3 has a calculated <a>, of about 0.28 mi?. The total area is
reapportioned using Table A2e-4 to assign 0.11 mi? to the FO classification, 0.13 mi? to the F1
classification, and 0.04 mi? to the F2 classification.

The risk regionalization scheme from NUREG/CR-2944, as shown in Figure A2e-4, was used
to determine the exceedance probability for each region identified. A continental U.S. average
was also determined. Figure A2e-4 shows the approximate location of commercial LWRs and
independent spent fuel storage facilities.

The SPC raw data for each state was used to determine the F-scale, path-length and path-
width characteristics of the reported tornadoes. The effective tornado strike area was corrected
using the data from NUREG/CR-2944. Equation A2e-1 was used for each state and the
summation and averaging of the states within each region (A, B, C and D, as well as a
continental USA average) performed. The results for the exceedance probability per year for
each F-scale are given in Table A2e-5, and graphically presented in Figure A2e-5. The SPC
data analysis is summarized in Table A2e-6.

4. SIGNIFICANT POOL DAMAGE

An F4 to F5 tornado would be needed to consider the possibility of damage to the SFP by a
tornado missile. The likelihood of having or exceeding this size tornado is estimated to be
5.6x107 per year (for Region A), or lower. In addition, the SFP is a multiple-foot thick concrete
structure. Based on the DOE-DOE-STD-1020-94 information, it is very unlikely that a tornado
missile would penetrate the SFP, even if it were hit by a missile generated by an F4 or F5
tornado.

Appendix 2E A2E-3


accel above 1.2
is<4.5E6

Changes may be implemented under
Meet seismic checklist revised rule |

‘Meet IDCs and SDAs
i

Plant-specific internal

events i
i}
a A
+
> submit risk analysis for staff review

Total freq <IE-S/y and other
RG 1.174 criteria are met

Changes may be implemented via |
exemption to existing rule

|

Appendix 4C. A4C-6


from INEL-96/0334. These individual pump failures result in a value of 0.18 for event
FP-DGPUMP-FTF and 6.7E-4 for event FP-2PUMPS-FTF.

The dependency between makeup water supply (e.g., fragility of the fire water supply tank) to
events that may have caused the loss of offsite power (such as high winds) is assumed to be
bounded by the dependency modeled in the HEPs.

4.4.4.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-RECG-FWST-SW 1.0E-4
HEP-RECG-DEPEN 5.0E-2
HEP-FW-START-SW 1.0E-3
HEP-FW-REP-DEPSW 7.0E-2
HEP-FW-REP-NODSW 1.8E-2
FP-2PUMPS-FTF 6.7E-4
FP-DGPUMP-FTF 1.8E-1

4.4.5 Top Event OFD - Operator Recovery Using Offsite Sources
4.4.5.1 Event Description and Timing

Given the failure of recovery actions using onsite sources, this event accounts for recovery of
coolant makeup using offsite sources such as procurement of a fire engine. Adequate time is
available for this action, provided that the operator recognizes that recovery of cooling using
onsite sources will not be successful, and that offsite sources are the only viable alternatives.
Fault tree LP2-OFD represents this top event for the lower branch (offsite power not recovered),
and LP2-OFD-U for the upper branch. These fault trees contain those basic events from the
fault trees LP2-OMK-U and LP2-OMK-L that relate to recognition of the need to initiate the
firewater system; if OMK fails because the operator failed to recognize the need for firewater
makeup, then it is assumed that the operator will fail here for the same reason.

4.4.5.2 Relevant Assumptions
¢ The operators have 88 hours to provide makeup and inventory cooling.
e Procedures and training are in place that ensure that offsite resources can be brought to

bear (NEI commitment no. 2, 3 and 4), and that preparation for this contingency is made
when it is realized that it may be necessary to supplement the poo! make-up.

Appendix 2A A2A-49


APPENDIX 6
STAKEHOLDER CONCERNS RAISED DURING THE PUBLIC COMMENT PERIOD

On February 15, 2000, the Nuclear Regulatory Commission (NRC) released the "Draft Final
Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Plants," for public
comment. The NRC encouraged stakeholders to review the draft study and to formally submit
comments for review. Appendix 7 of that report included a list of public meetings and how the
staff addressed stakeholder comments received on the draft report, issued June 1999, in
various technical areas. After review of the February 2000 study, several public groups
commented that it appeared that the NRC did not address some of the public’s comments.
While all stakeholder comments were considered and many resulted in changes in the study,
the staff did not include a discussion for some of the comments in Appendix 7. In order to
ensure that adequate consideration had been given to public comments, the staff reviewed
comments which had been received prior to February 15, 2000, as well as comments received
as a result of a review of the draft final report. Comments received prior to February 15 were
identified by reviewing transcripts of publically attended meetings, letters from the public, and
other available documentation related to the staff's efforts in completing the draft final report.

This appendix provides the NRC’s responses to the comments and concerns received as
described above. In most cases, responses are documented in this appendix. However, in
other cases, comments or concerns identified in this appendix are referred to other parts of the
report where the identified issues are addressed. For cases where similar comments were
received by more than one commenters, the comments were combined for one response. The
comments are grouped in the following technical categories: Criticality, Consequences,
Probability and Human Reliability, Seismic, Security/Safety Culture/EP, Thermal hydraulics,
Insurance, and Rulemaking/ NRC Process Concerns.

CRITICALITY
Comment #1: A commenter stated that the potential criticality should be addressed.

Response: The staff agrees. The issue of nuclear criticality is addressed in Section 3.6 and
Appendix 3.

Comments #2 and 3: A commenter raised several concerns related to SFP criticality. (a) Can
a criticality occur due to chemical stripping of primary piping? (b) During primary system
decontamination at decommissioning reactors, is it possible to misalign the valves and send
corrosive chemicals into the SFP? Could these chemicals precipitate boron from the SFP
water? Is there a potential for criticality? Is there a potential for fuel damage?

Response: The precipitation of boron out of the pool water, due to chemicals or any other
means, will not result in criticality because soluble boron is not credited to maintain spent fuel
pool subcriticality (k-eff < 1.0). The main connection between the spent fuel pool and primary
system is the transfer tube used to transfer fuel for refueling. After a plant ceases to operate,
this tube is sealed on both ends with flanges. As a result, there is no communication between
the primary system and SFP from this connection. Support systems connected to the SFP vary
from one plant to another. At most decommissioning plants, there would be no communication
between the SFP and the primary reactor systems, while others may use a primary support
water system to add water to the pool. In any event, even if fuel damage did occur, the
shielding provided by the large volume of water above the fuel (usually 23 feet of water) would

Appendix 6 A6-1


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3-32


Table A2e-1 Tornado Characteristics

Path-length scale Path-width scale
F-scale Damage and wind speed
Scale Length (mi) Scale Width (yds)
0 Light Damage (40-72 mph) | 0 <1.0 0 <18
1 Moderate Damage (73-112 mph) | 1 1.0-3.1 1 18-55
2 Significant Damage (113-157 mph) 2 3.2-9.9 2 56-175
3 Severe Damage (158-206 mph) 3 10.0 - 31.9 3 176 - 527
4 Devastating Damage (207-260 mph) 4 32 - 99.9 4 528 - 1759
5 Incredible Damage (261-318 mph) 5 100 > 5 1760 >
Table A2e-2 Variation of Intensity Along Length
Based on Fraction of Length per Tornado”
Local Recorded tomado state
tomado
state FO FA F2 F3 F4 FS
PL-FO 4 0.383 0.180 0.077 0.130 0.118
PL-F1 0.617 0.279 0.245 0.131 0.125
PL-F2 0.541 0.310 0.248 0.162
PL-F3 0.368 0.234 0.236
PL-F4 0.257 0.187
PL-FS 0.172

(*) - Table 6b from NUREG/CR-2944

Table A2e-3 Variation of Intensity Alo:

ng Width Based on Fraction of Width Per Tornado”

Local Recorded tornado state
tomado

state FO FA F2 F3 F4 FS
PW-FO 1 0.418 0.154 0.153 0.152 0.152
PW-F1 0.582 0.570 0.310 0.264 0.262
PW-F2 0.276 0.363 0.216 0.143
PW-F3 0.174 0.246 0.168
PW-F4 0.122 0.183
PW-F5 0.092

(*) - Table 7b from NUREG/CR-2944

Appendix 2E

A2E-5



for the diaphragm compared to the same amount of damage in a wall.

Based upon the guidance provided in Reference 1, diaphragms for Category | structures
designed for a SSE of 0.1g or greater do not require an explicit evaluation provided that: (1) the
diaphragm loads were developed using dynamic analysis methods; (2) they comply with the
ductility detailing requirements of ACI 318-71 or ACI 349-76 or later editions. Diaphragms which
do not comply with the above ductility detailing or which did not have loads explicitly calculated
using dynamic analysis should be evaluated for a beyond-design-basis seismic event in the
0.45-0.5g pga range.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 4:

Requirement: Verify the Adequacy of the SFP Walls and Floor Slab to Resist Out-of-

Plane Shear and Flexural Loads

Basis: For PWR pools that are fully or partially embedded, an earthquake motion that could
cause a catastrophic out-of-plane shear or flexural failure is very high and is not a credible
event. For BWR pools (and PWR pools that are not at least partially embedded), the seismic
capacity is likely to be somewhat less and the potential for out-of-plane shear and/or flexural
wall or base slab failure, at beyond-design-basis seismic loadings, is possible.

A structural assessment of the pool walls and floor slab out-of plane shear and flexural
capabilities should be performed and compared to the realistic loads expected to be generated
by a seismic event equal to approximately three times the site SSE. This assessment should
include dead loads resulting from the masses of the pool water and racks, seismic inertial
forces, sloshing effects and any significant impact forces.

Credit for out-of-plane shear or flexural ductility should not be taken unless the
reinforcement associated with each failure mode can be shown to meet the ACI 318-71 or ACI
349-49 requirements.

Design Feature: Compliance with this design feature will be documented basedupon a
review of drawings (in the case of embedded or partially embedded PWR pools) or based
upon a review of drawings coupled with the specified beyond-design-basis shear and
flexural calculations outlined above.

Item 5:
Requirement: Verify the Adequacy of Structural Steel (and Concrete) Frame Construction

Basis: At a number of older nuclear power plants, the walls and roof above the top of the spent
fuel pool are constructed of structural steel. These steel frames were generally designed to
resist hurricane and tornado wind loads which exceeded the anticipated design basis seismic
loads. A review of these steel (or possibly concrete) framed structures should be performed to
assure that they can resist the seismic forces resulting from a beyond-design-basis seismic
event in the 0.45-0.5g pga range. Such a review of steel structures should concentrate on

A2B-8


APPENDIX 2C
STRUCTURAL INTEGRITY OF SPENT FUEL POOL STRUCTURES SUBJECT TO
HEAVY LOADS DROPS

1. INTRODUCTION

A heavy load drop into the spent fuel pool (SFP) or onto the SFP wall can affect the structural
integrity of the SFP. A loss-of-inventory from the SFP could occur as a result of a heavy load
drop. For single failure-proof systems where load drop analyses have not been performed at
decommissioning plants, the mean frequency of a loss-of-inventory caused by a cask drop was
estimated to be 2.0x107 per year (assuming 100 lifts per year). For a non-single failure-proof
handling system where a load drop analysis has not been performed, the mean frequency of a
loss-of-inventory event caused by a cask drop was estimated to be 2.1x10° per year. The staff
believes that performance and implementation of a load drop analysis that has been reviewed
and approved by the staff will substantially reduce the expected frequency of a loss-of-inventory
event from a heavy load drop for either a single failure-proof or non-single failure-proof system.

2. ANALYSIS

The staff revisited NUREG-0612? (Ref. 1) to review the evaluation and the supporting data
available at that time to determine its applicability to and usefulness for evaluation of heavy load
drop concerns at decommissioning plants. In addition, three additional sources of information
were identified by the staff and used to reassess the heavy load drop risk:

(1) U.S. Navy crane experiences (1990s Navy data) for the period 1996 through mid-1999.

(2) WIPP/WID-96-2196 (Ref. 2), “Waste Isolation Pilot Plant Trudock Crane System Analysis,”
October 1996 (WIPP).

(3) NEI data on actual SFP cask lifts at U.S. commercial nuclear power plants (Ref.3).

The staff's first area of evaluation was the frequency of heavy load drops. The number of
occasions (incidents) where various types of faults occurred that potentially could lead to a load
drop was investigated. Potential types of faults investigated included improper operation of
equipment, improper rigging practices, poor procedures, and equipment failures. Navy data
from the 1990s were compared to the data used in NUREG-0612. The data gave similar, but
not identical, estimates of the various faults leading to heavy load drops (see Table A2c-1). The
NEI cask handling experience also supported the incident data used in this evaluation, and in
NUREG-0612. Once the frequency of heavy load drops was estimated (i.e., load drops per lift),
the staff investigated the conditional probability that such a drop would seriously damage the
SFP (either the bottom or walls of the pool) to the extent that the pool would drain very rapidly
and it would not be possible to refill it using onsite or offsite resources. To do this the staff used
fault trees taken from NUREG-0612 (see Figure A2c-1). By mathematically combining the

?NUREG-0612 documented the results of the staff's review of the handling of heavy
loads at operating nuclear power plants and included the-staff's recommendations on actions
that should be taken to assure safe handling of heavy loads.

Appendix 2C. A2C-1


generic basis. Other policy considerations beyond the scope of this technical study will
therefore need to be considered for insurance requirements.

The study also discusses implications for security provisions at decommissioning plants. For
security, risk insights can be used to determine what targets are important to protect against
sabotage. However, any revisions in security provisions should be constrained.by an
effectiveness assessment of the safeguards provisions against a design-basis threat. Because
the possibility of a zirconium fire leading to a large fission product release cannot be ruled out
even many years after final shutdown, the safeguards provisions at decommissioning plants
should undergo further review. The results of this study may have implications on previous
exemptions at decommissioning sites, devitalization of spent fuel pools at operating reactors
and related regulatory activities.

The staff's risk analyses were complicated by a lack of data on severe-earthquake return
frequencies, source term generation in an air environment, and SFP design variability. Although
the staff believes that decommissioning rulemaking can proceed on the basis of the current
assessment, more research may be useful to reduce uncertainties and to provide insights on
operating reactor safety. In particular, the staff believes that research may be useful on source
term generation in air, which could also be important to the risk of accidents at operating
reactors during shutdowns, when the reactor coolant system and the primary containment may
both be open.

In summary, the study finds that:

1. The risk at decommissioning plants is low and well within the Commission’s safety goals.
The risk is low because of the very low likelihood of a zirconium fire even though the
consequences from a zirconium fire could be serious.

2. — The overall low risk in conjunction with important differences in dominant sequences
relative to operating reactors, results in a small change in risk at decommissioning plants if
offsite emergency planning is relaxed. The change is consistent with staff guidelines for
small increases in risk.

3. Insurance, security, and emergency planning requirement revisions need to be considered
in light of other policy considerations, because a criterion of “sufficient cooling to preclude
a fire” cannot be satisfied on a generic basis.

4. Research on source term generation in an air environment would be useful for reducing
uncertainties.

xi


Table A2d-1 Generic Aircraft Data

Aircraft Wingspan Skid distance cote Crashes per mi*/yr Notes
® ® Min Ave Max
General aviation 50 1440 10.2 1x107 2x10* 3x10°
Air carrier 98 60 8.2 7x10 4x107 2x10®
Air taxi 58 60 8.2 4x107 1x10* 8x10°
Large military 223 780 74 6x10 2x107 7x107 takeoff
Small military 100 447 10.4 4x10* 4x10*® 6x10 landing
Table A2d-2 Aircraft Hits Per Year
Building (L x Wx H) Average Minimum hits Average hits Maximum hits
(ft) effective area (mi?) (per year) (per year) (per year)
With the DOE effective aircraft
target area model
100 x 50 x 30 6.9x10° 3.2x10° 2.1x10® 3.1x105
200 x 100 x 30 1.1x10? 5.3x10° 3.7x10* 5.5x10%
400 x 200 x 30 2.1x107 1.0x10° 7.0x10* 1.0x10*
200 x 100 x 100 1.8x10? 9.6x10° 5.1x10* 7.6x10*
400 x 200 x 100 3.3x10? 1.8x10°% 9.6x10* 1.4x10*
80 x 40x 30 6.1x10% 2.8x10° 1.8x10* 2.710%
10 x 10x 10 2.9x10°% 1.1x10° 7.3x107 1.1x10%
With the point target area
model
100 x 50 xO 1.8x104 1.2x10°7° 3.7x10* 5.4x107
200 x 100 x 0 7.2x104 4.8x107° 1.5x107 2.2x10°
400 x 200 x 0 2.9x10% 1.9x10° 5.9x107 8.6x10*
80x 40x0 1.1x10* 4.1x10" 2.4x10® 3.5x107
10x 10 3.6x10° 2.4x10°7? 7.4x10°° 1.1x10®
Appendix 2D A2D-5



neutron-absorbing material. If BWR pools were borated, criticality would not be possible
during a low-density rack compression event.

(2) If the stored assemblies are separated by neutron absorber plates (e.g., Boral or Boraflex),
loss of these plates could result in a potential for criticality for BWR pools. For PWR pools,
the soluble boron in the fuel pool water is sufficient to maintain subcriticality. The absorber
plates are generally enclosed by cover plates (stainless steel or aluminum alloy). The
tolerances of cover plates tend to prevent any appreciable fragmentation and movement of
the enclosed absorber material. The total loss of the welded cover plate is not considered
feasible.

Boraflex has been found to degrade in spent fuel pools because of gamma radiation and
exposure to the wet pool environment. For this reason, the NRC issued Generic

Letter 96-04 on Boraflex degradation in spent fuel storage racks to all holders of operating
licenses. Each addressee that uses Boraflex was requested to assess the capability of the
Boraflex to maintain a 5-percent subcriticality margin and to submit to the NRC proposed
actions to monitor the margin or confirm that this 5 percent margin can be maintained for
the lifetime of the storage racks. Many licensees subsequently replaced the Boraflex racks
in their pools or reanalyzed the criticality aspects of their pools, assuming no reactivity
credit for Boraflex.

Other potential criticality events, such as events involving loose pellets or the impact of water
(adding neutron moderation) during personnel actions in response to accidents, were
discounted because the basic physics and neutronic properties of the racks and fuel would
prevent criticality conditions from being reached with any credible likelihood. For example,
without moderation fuel at current enrichment limits (no greater than 5 wt% U-235) cannot
achieve criticality, no matter what the configuration. If it is assumed that the pool water is lost, a
reflooding of the storage racks with unborated water may occur during personnel actions.
However, both PWR and BWR storage racks are designed to remain subcritical if moderated by
unborated water in the normal configuration. Thus, the only potential credible scenarios are the
two scenarios described above, which involve crushing of fuel assemblies in low-density racks
or degradation of Boraflex over long periods in time. These conclusions assume present light-
water uranium oxide reactor fuel designs. Alternative fuel designs, such as mixed oxide (MOX)
fuels will have to be reassessed to ensure that additional vulnerabilities for pool criticality do not
exist.

To gain qualitative insights on credible criticality events, the staff considered the sequences of
events that must occur. For scenario 1, a heavy load drop into a low-density racked BWR pool,
compressing the assemblies would be required. From its analysis of the heavy load drop
documented in Appendix 2C, the staff has determined the likelihood of a heavy load drop from a
single failure-proof crane to have a mean frequency of approximately 9.6X10° per year,
assuming 100 cask movements per year at the decommissioning facility. From the load path
analysis done in Appendix 2C, the staff estimates that the load is over or near the pool
approximately 13 percent of the movement path length, depending on the plant's layout. The
additional frequency reduction in the appendix to account for the fraction of time that the heavy
load is lifted high enough to damage the pool liner is not applicable here because the fuel
assemblies can be crushed by a smaller impact velocity than required to need to crush the pool

3-26


~2-

control room (or where personnel are stationed) for spent fuel pool
temperature, water level, and area radiation levels.

6. Spent fuel pool boundary seals that could cause leakage leading to fuel
uncovery in the event of seal failure shall be self limiting to leakage or
otherwise engineered so that drainage cannot occur.

7. Procedures or administrative controls to reduce the likelihood of rapid
drain down events will include (1) prohibitions on the use of pumps that
lack adequate siphon protection or (2) controls for pump suction and
discharge points. The functionality of anti-siphon devices will be
periodically verified.

8. Anon site restoration plan will be in place to provide repair of the spent
fuel pool cooling systems or to provide access for make-up water to the
spent fuel pool. The plan will provide for remote alignment of the make-up
source to the spent fuel pool without requiring entry to the refuel floor.

9. Procedures will be in place to control spent fuel pool operations that have
the potential to rapidly decrease spent fuel pool inventory. These
administrative controls may require additional operations or management
review, management physical presence for designated operations or
administrative limitations such as restrictions on heavy load movements.

10. Routine testing of the alternative fuel pool make-up system components
will be performed and administrative controls for equipment out of service
will be implemented to provide added assurance that the components
would be available, if needed.

If you have any questions regarding industry's commitments, please contact me at
202 739-8109 or LXII@NEI.org.

Sincerely,

Lynnette Hendricks
LXH/1rh


Figure 4.1

Loss of SFP cooling system event tree

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1 coouns ROOM | INDICATIONS | RECOVERY | INTIATES | USING |
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Appendix 2A A2A-16


3. Development and Use of Seismic Screening Criteria

Screening criteria are very useful to reduce the number of structure, system, and
component (SSC) failure modes for which either seismic fragilities or seismic margin HCLPF
capacities need to be developed. Screening criteria are presented in Ref. 6 for SSCs for which
failures might lead to core damage. These screening criteria were established by an NRC
sponsored "Expert Panel” based upon their review of seismic fragilities and seismic margin
HCLPF capacities computed for these SSCs at more than a dozen nuclear power plants, and their
review of earthquake experience data. These screening criteria were further refined in Ref. 7.

The screening criteria of Refs. 6 and 7 are defined for two seismic margin HCLPF
capacity levels which will be herein called Level 1 and Level 2. Refs. 6 defines these two
HCLPF capacity levels in terms of the PGA of the ground motion. However, damage to critical
SSCs does not correlate very well to PGA of the ground motion. Damage correlates much better
with the spectral acceleration of the ground motion over the natural frequency range of interest
which is generally between 2.5 and 10 Hz for nuclear power plant SSCs. For this reason, Ref. 7
defines these same two HCLPF capacity levels in terms of the peak 5% damped spectral
acceleration (PSA) of the ground motion. The two HCLPF capacity screening levels defined in
Refs 6 and 7 are:

HCLPF Screening Levels
Level 1 Level 2
PGA (Ref. 6) 0.3g 0.5g
PSA (Ref. 7) 0.8g 1.2g

These two definitions (PGA and PSA) are consistent with each other based upon the data
upon which these screening levels are based. However, in my judgment, it is far superior to use
the Ref. 7 PSA definition for the two screening levels when convolving a fragility estimate with
CEUS seismic hazard estimates. For these CEUS seismic hazard estimates from Ref. 8, the ratio
PSA/PGA generally lies in the range of 1.8 to 2.4 which is lower than the PSA/PGA ratio of the
data from which the screening tables were developed. A more realistic and generally lower
estimate of the annual probability of failure will result when the seismic fragility is defined in
terms of PSA and convolved with a PSA hazard estimate in which the PSA hazard estimate is
defined in the 2.5 to 10 Hz range.

In the past, a practical difficulty existed with defining the seismic fragility in terms of
PSA instead of PGA. The Ref. 8 PSA hazard estimates are only carried down to 10“ annual
frequency of exceedance whereas the PGA hazard estimates are extended down to about 10°.
Since it is necessary for the hazard estimate to be extended to at least a factor of 10 below the
annual failure frequency being predicted, it has not been practical to use the PSA seismic
fragility definition with the Ref. 8 hazard estimates. However, this difficulty has been overcome
by Ref. 9 prepared by the Engineering Research Applications Branch of the Nuclear Regulatory

A2B-14


+ For the larger loss-of-coolant-inventory accidents, water addition through the makeup
pumps does not successfully mitigate the loss of the inventory event unless the location of
inventory loss is isolated.

+ The SFP fuel handlers perform walkdowns of the SFP area once per shift (8- to 12-hour
shifts). A different crew member works the next shift. The SFP water is clear and the pool
level is observable via a measuring stick in the pool to alert fuel handlers to level changes.

+ Plants do not have drain paths in their SFPs that could lower the pool level (by draining,
suction, or pumping) more that 15 feet below the normal pool operating level, and licensees
must initiate recovery using offsite sources.

Based upon the results of the June 1999 preliminary risk analysis and the associated sensitivity
cases, it became clear that many of the risk sequences were quite sensitive to the performance
of the SFP operating staff in identifying and responding to off-normal conditions. This is
because the remaining systems of the SFP are relatively simple, with manual rather than
automatic initiation of backups or realignments. Therefore, in scenarios such as loss of cooling
or inventory loss, the fuel handler’s responses to diagnose the failures and bring any available
resources (public or private) to bear is fundamental for ensuring that the fuel assemblies remain
cooled and a zirconium fire is prevented.

As part of its technical evaluations, the staff assembled a small panel of experts‘ to identify the
attributes necessary to achieving very high levels of human reliability for responding to potential
accident scenarios in a decommissioning plant SFP. (These attributes and the human reliability
analysis (HRA) methodology used are discussed in Section 3.2 of Appendix 2A.)

Upon considering the sensitivities identified in the staff's preliminary study and to reflect actual
operating practices at decommissioning facilities, the nuclear industry, through NEI, made
important commitments, which are reflected in the staff's updated risk assessment.

Industry Decommissioning Commitments (IDCs)

IDC #1 Cask drop analyses will be performed or single failure-proof cranes will be in use for
handling of heavy loads (i.e., phase Il of NUREG-0612 will be implemented).

IDC #2 Procedures and training of personnel will be in place to ensure that onsite and offsite
resources can be brought to bear during an event.

IDC #3 Procedures will be in place to establish communication between onsite and offsite
organizations during severe weather and seismic events.

‘Gareth Parry, U.S. NRC; Harold Blackman, INEEL; and Dennis Bley, Buttonwood
Consulting.

3-5


methods were applied to determine each parameter, and sensitivity studies were performed to
account for the uncertainties in the geophysical phenomena. In addition, nuclear power plants
have design margins (capability) well beyond the demands of the SSE. The ability of a nuclear
power plant to resist the forces generated by the ground motion during an earthquake is
thoroughly incorporated in the design and construction. As a result, nuclear power plants are
able to resist earthquake ground motions beyond their design basis and above the ground
motion that would result in severe damage to residential and commercial buildings designed
and built to standard building codes.

Following large damaging earthquakes such as the Kobe and Northridge events, the staff
reviewed the seismological and engineering information obtained from these events to
determine if the new information challenged previous design and licensing decisions. The
Kobe and Northridge earthquakes were tectonic plate boundary events occurring in regions of
very active tectonics. The operating U.S. nuclear power plants (except for San Onofre and
Diablo Canyon) are located in the stable interior portion of the North American tectonic plate.
This is a region of relatively low seismicity and seismic hazard. Earthquakes with the
characteristics of the Kobe and Northridge events will not occur near central and eastern U.S.
nuclear power plant sites.

The ground motion from an earthquake at a particular site is a function of the earthquake
source characteristics, the magnitude and the focal mechanism. It is also a function of the
distance of the facility to the fault, the geology along the travel path of the seismic waves, and
the geology immediately under the facility site. Two U.S. operating nuclear power plant sites
can be considered as having the potential to be subjected to the near field ground motion of
moderate to large earthquakes. These are the San Onofre Nuclear Generating Station
(SONGS) near San Clemente and the Diablo Canyon Power Plant (DCPP) near San Luis
Obispo. The seismic design of SONGS Units 2 and 3 is based on the assumed occurrence of
a Magnitude 7 earthquake on the Offshore Zone of Deformation, a fault zone approximately

8 kilometers from the site. The design of DCPP has been analyzed for the postulated
occurrence of a Magnitude 7.5 earthquake on the Hosgri Fault Zone, approximately

4 kilometers from the site. The response spectra, used for both the SONGS and the DCPP,
was evaluated against the actual spectra of near field ground motions of a suite of earthquakes
gathered on a worldwide basis.

The commenter stated, “... it's clear that new information was gained out of Kobe and
Northridge events suggesting that you can have seismological effects of greater consequence
farther afield than at the epicenter of the event.” A review of the strong motion data and the
damage resulting from these events does not bear out the validity of this concern at SONGS
and DCPP.

The staff assumes that the individual alluded to the fact that the amplitudes of the ground
motion from the 1994 Northridge earthquake were larger in Santa Monica than those at similar
and lesser distances from the earthquake source. The cause of the larger ground motions in
the Santa Monica area is believed to be the subsurface geology along the travel path of the
waves. One theory (Gao et al, 1996) is that the anomalous ground motion in Santa Monica is
explained by focusing due to a deep convex structure (several kilometers beneath the surface)
that focuses the ground motion in mid-Santa Monica. Another theory (Graves and Pitarka,
1998) is that the large amplitudes of the ground motions in Santa Monica from the Northridge
earthquake are caused by the shallow basin-edge structure (1 kilometer deep) at the northern
edge of the Los Angles Basin. This theory suggests that the large amplification results from
constructive interference of direct waves with the basin-edge generated surface waves.
Earthquake recordings at San Onofre and Diablo Canyon do not indicate anomalous

Appendix 6 A6-19


Heatup Time to Release (Air Cooling)

ay
Oo

| Z

w
=)

LZ

PWR

=
oO
—t

BWR

heatup time (hours)
N
(=)

yA

ote 7 3 3 4
shutdown time (years)
Figure 2.1 Heatup Time From 30 °C to 900 °C

PWR Adiabatic vs. Air cooled

w
oO

2

Ny oN
a
Ni

\|

heatup time (hours)
So a

i

0.164. 4 3 3 4 5

shutdown time (years)

Figure 2.2 PWR Heatup Times for Air Cooling and Adiabatic Heatup

2-3

=

Air Cooled

——

Adiabatic


Table 4.1-1 Industry Decommissioning Commitments
Table 4.1-2 Staff Decommissioning Assumptions ..................00000 cece eee

Appendices

Appendix 1A = Thermal-hydraulics Analysis of Spent Fuel Pool Heatup.............
Appendix 1B Temperature Criteria for Spent Fuel Pool Analysis ............. :
Appendix 2 Assessment of Spent Fuel Pool Risk at Decommissioning Plants
Appendix 2A _— Detailed Assessment of Risk from Decommissioning Plant Spent Fuel
POOlS 2.2.1. e cee eee eee eect eee nets A2A-1
Appendix 2B — Structural Integrity of Spent Fuel Pools Subject to Seismic Loads ..... A2B-1
Enhanced Seismic Checklist
Comments Concerning Seismic Screening And Seismic Risk of Spent
Fuel Pools for Decommissioning Plants by Robert P. Kennedy,
October 1999
Response to Questions Concerning Spent Fuel Pool Seismic-Induced
Failure Modes and Locations and the Expected Level of Collateral

Damage
Appendix 2C — Structural Integrity of Spent Fuel Pool Structures Subject to Heavy Loads
Drops .. 2... eee ete tee etree eens A2C-1
Appendix 2D _ Structural Integrity of Spent Fuel Pool Structures Subject to Aircraft
CrasheS eee eee teen eee A2D-1

Appendix 2E —_ Structural Integrity of Spent Fuel Pool Structures Subject to Tornados . A2E-1
Appendix 3 Assessment of the Potential for Criticality in Decommissioning Spent Fuel

Pool
Appendix 4 Consequence Assessment from Zirconium Fire . .
Appendix 4A __ Risk-informed Requirements for Decommissioning................. A4A-1
Appendix 4B —_—- Radiological Consequences of Spent Fuel Pool Accident Occurring Up to

10 Years After Final Reactor Shutdown ...................2.0000. A4B-1

Appendix 4C — Pool Performance Guideline ............... 00.0... cee eee eee
Appendix 4D Change in Risk Associated with EP Relaxations . .

Appendix 5 November 12, 1999 Nuclear Energy Institute Commitment Letter oe
Appendix 6 Public Comments ............. 0020 ccc eee ene e eee eee

vii


4.2.3.3 Quantification

Human Error Probability

This event is represented by the basic event HEP-WLKDWN-LSFPC which models the
operators’ failure to recognize the loss of cooling during walk-downs. The failure rate was
developed using THERP, and is based upon three individual failures: failure to carry out an
inspection, missing a step in a written procedure, and misreading a measuring device. Multiple
opportunities for recovery were assumed.

Note that no dependency on the previous HEP was modeled. While it could be argued that, in
the case where the operator has already failed to respond to control room alarms, there may be
a dependence between the event HEP-DIAG-ALARM and HEP-WLKDWN-LSFPC. However,
the cues for this event are quite different. There will be obvious physical changes in the plant
(e.g., loss of offsite power, a burnt out area, smoke, etc.). The only source of dependency is
one where a situation would result in the operators failing to respond to control room alarms and
also result in a total abandonment of plant walk-downs.

4.2.3.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-WLKDWN-LSFPC. 1.0E-5

4.2.4 Top Event OSP - Fire Suppression

4.2.4.1 Event Description and Timing

This top event represents operator failure to suppress the fire before the SFP cooling system is
damaged given that he responds to fire alarms. If the SFP cooling and makeup system pumps
and plant power supply system are damaged to a point that they cannot be repaired in time to
prevent fuel uncovery, the operator must provide cooling using available onsite (i.e., diesel fire
pumps) and offsite water sources. If the fire is suppressed in time to prevent damage to SFP
components, then the SFP cooling system can be restored in time to prevent fuel uncovery.
The top event is represented by fault tree FIR-OSP.

4.2.4.2 Relevant Assumptions

. The automatic fire suppression system is unavailable.

. If the fire is not extinguished within 20 minutes, it is assumed that SFP cooling will be lost
due either to damage of SFPC equipment, or to the plant's power supply system.

° No credit is taken for the firewater system in the suppression of the fire.
. Fire suppression extinguishers are located strategically in the SFP area, and these

extinguishers are tested periodically.

Appendix 2A A2A-28


Support Document for the Aircraft Crash Risk Analysis Methodology (ACRAM) Standard,”

C.Y. Kimura, et al., UCRL-ID-124837, Lawrence Livermore National Laboratory, August 1,
1996). The number of U.S. air carrier operations increased from about 5.5 million departures
per year in the 1970s to about 8.7 million departures per year in the mid-1990s. The average
miles traveled per departure increased from about 500 to 650. For the period from 1986 to
1993 general aviation operations remained relatively constant, with a decrease in activities
reported in 1992 and 1993. Military aircraft data, which are a small fraction of the total risk (see
Appendix 2D, Table A2d-1, “Generic Aircraft Data’), was not reviewed.

Comments #18 and 19: (A) What is the generic frequency of events leading to zirconium fires
at decommissioning plants before the implementation of industry commitments and staff
assumptions? (B) This question is relevant to operating plants.

Response: Risk assessments are performed as realistic as possible. As such, the analysis for
this study reflects practices already in place. The staff visited four decommissioning sites as
part of the preparation for developing the risk assessment of decommissioning spent fuel pools.
The insights from those visits include that the facilities appeared to have been staffed by well
trained and knowledgeable individuals with significant nuclear power plant experience.
Procedures were in place for dealing with routine losses of inventory. Fuel handlers appeared
to know whom to contact off-site if difficulties arose with the SFP. The staff recognized that
these attributes were not required by NRC regulations nor suggested in NRC guidance for
decommissioning sites. The IDCs and SDAs are an attempt to increase the assurance that
plant personnel will continue to be knowledgeable of off-site resources and have good
procedures available to them.

This study does not reflect the risk at operating plants. As with the practices discussed above,
this study reflects the support systems and staffing generally found at decommissioning plants,
which are different than at operating plants. For example, the spent fuel pool cooling and
makeup systems at decommissioning plants are generally replacing smaller capacity systems
to match the reduced decay heat level of the spent fuel. The staff believes that a direct
comparison of this risk study on decommissioning plants can not be made to operating plants.
However, the staff is sensitive to possible implications to operating plants.

Comment #20: There are several places in the draft report where the staff refers to
“uncovering the core” rather than “uncovering the fuel.”

Response: The phrase “uncovering the core” has been replaced by “uncovering the fuel.”

Comment #21: Recalculating the frequencies for event trees produced numerical results for
some sequences that were off by one or two orders of magnitude.

Response: In the staff's risk analysis, the accident scenario frequencies in the event trees
were calculated such that dependencies among the failure events (in the event tree branches)
were taken into account. Therefore, if an event resulted in functional failure in more than one
branch in the event tree, this dependency was taken into account, and the resultant scenario
frequency is therefore larger (in some cases, by as much as two orders of magnitude) than if
the events were assumed to be independent.

Appendix 6 A6-7


The Base Case calculation assumed that evacuation begins about an hour after the fission product
release begins. However, Appendix 1 states that, after a year of decay, it will take a number of hours for
the fuel with the highest decay power density to heat up to the point of releasing fission products in the
fastest progressing accident scenarios. As a result, it is more likely to have evacuation before the
release begins. Therefore, the Base Case calculation then was modified to begin the evacuation three
hours before the fission product release begins. This modified Base Case is called Case 13. Starting
with Case 13, sensitivity cases were run with a ruthenium release fraction of one and a uniform
population density of 100 people/mile®. The results of these cases (i.e., Cases 13, 14, 15, 16) are given
in Table 3. For these cases, the effect of ruthenium is to increase the number of prompt fatalities by a
factor of six to 30. The effect on societal dose and cancer fatalities is a more modest increase, with the
largest effect being a factor-of-three increase in cancer fatalities for the Surry population density.

For the cases in Table 3, the total number of prompt fatalities increases by a larger factor for Surry than
for the uniform population density when a significant ruthenium release is included. Therefore, as part of
the ruthenium sensitivity assessment, the staff further examined the effect of population density on
prompt fatalities. For the cases with late evacuation (i.e., Base Case, Cases 11, 21, 22), Table 4 gives
the MACCS results for the individual risk of a prompt fatality in each radial ring which is composed of 16
sectors. The individual risk of a prompt fatality is a function of the dose to an individual and is
independent of the population density. The total number of prompt fatalities is calculated in MACCS by
multiplying, in each sector, the individual risk of a prompt fatality by the total number of people in that
sector. Table 5, which is the result of multiplying the individual risk of a prompt fatality in each ring from
Table 4 by the population in each ring, indicates that Surry’s higher increase in prompt fatalities is
caused by the jump in the Surry population density at 8.1 km shown in Table 4.

Table 4 Individual Risk of a Prompt Fatality for Cases with Late Evacuation

Distance Individual risk of a prompt fatality Ratio Surry
(km) population
Base Case and Case 21, Cases 11 and 22, tbetecne!
Ru release fraction of 2x10° Ru release fraction of 1 km?)
O-.2 | .146 -169 1.16 0
2-5 | .0302 0657 2.18 i)
5-12 | .0138 0374 2.71 1,33
1.2-1.6 | .00828 .0301 3.64 1.13
1.6-2.1 | .00575 .0266 4.63 1.80
2.1-3.2 | .00326 .0216 6.63 1.58
3.2-4.0 | 00151 0146 9.67 7.15
40-48 | .00167 .0132 7.90 7.77
48-56 | .00171 0110 6.43 7.84
5.6-8.1 | .0000672 .0131 194.94 8.07
8.1-11.3 | .000000254 .00301 11850.39 117.80
11.3-16.1 |0 .0000225 NA 118.36
16.1-20.9 | 0 0 NA 83.75

*This data is from the MACCS input file SURSIT.INP.

-3-


. Plant personne! must be able to interpret the indications (identify the source of the
problem) and formulate a plan that would mitigate the situation (Situation Assessment and
Response Planning).

. Plant personnel must be able to perform the actions required to maintain cooling of and/or
add water to the SFP (Response Implementation).

In the following sections, factors that are relevant to determining effective operator responses
are discussed. While not minimizing the importance of such factors as the establishment of a
safety culture and effective intra-crew communication, the focus is on factors which can be
determined to be present on a relatively objective basis. A review of LERs associated with
human performance problems involved in response to loss of fuel pool cooling revealed a
variety of contributing factors, including crew inexperience, poor communication, and inadequate
administrative controls. In addition, there were some instances of design peculiarities that made
operator response more complex than necessary.

The factors discussed below were used to identify additional assumptions made in the analysis
that the staff considered would provide for an effective implementation of the NEI commitments.

3.2.3.1 Awareness/Detection of Deviant Conditions

There are two types of monitoring that can be expected to be used in alerting the plant staff to
deviant conditions: a) passive monitoring in which alarms and annunciators are used to alert
operators; b) active monitoring in which operators, on a routine basis, make observations to
detect off-normal behavior. In practice both would probably be used to some extent. The
amount of credit that can be assumed depends on the detailed design and application of the
monitoring scheme.

In assessing the effectiveness of alarms there are several factors that could be taken into
account, for example:

. alarms (including control room indications) are maintained and checked/calibrated on a
regular basis

. the instruments that activate instruments and alarms measure, as directly as possible, the
parameters they purport to measure

. alarm set-point is not too sensitive, so that there are few false alarms
. alarms cannot be permanently canceled without taking action to clear the signal
. alarms have multiple set-points corresponding to increasing degradation

. the importance of responding to the alarms is stressed in plant operating procedures and
training

. the existence of independent alarms that measure different primary parameters (e.g., level,

Appendix 2A A2A-6


+ Applicability of the staff's generic risk assessment to a specific facility is assured by
SDA #1.

With regard to SFP risks and risk increases associated with EP relaxations, the staff concludes:

-  AnSFP facility that conforms with the IDCs and SDAs would meet the QHOs by one to two
orders of magnitude shortly after shutdown and with greater margins at later times.

+ The risk increase associated with relaxations in EP requirements is very small, even under
assumptions that maximize the effectiveness of emergency preparedness in seismic events
(.e., assigning consequences for the “full EP” case based on early evacuation and
consequences for the “relaxed EP” case based on late evacuation), and the QHOs continue
to be met with adequate margin.

+ Continued conformance with IDCs and SDAs provides reasonable assurance that the SFP
risk and risk increases associated with regulatory changes would remain small.

4.1.2 Defense-in-Depth

RG 1.174 states that the proposed change should be consistent with the defense-in-depth
philosophy.

In accordance with the Commission white paper on risk-informed regulation (March 11, 1999),
“Defense-in-depth is an element of the NRC’s Safety Philosophy that employs successive
compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or
naturally caused event occurs at a nuclear facility. The defense-in-depth philosophy ensures
that safety will not be wholly dependent on any single element of the design, construction,
maintenance, or operation of a nuclear facility. The net effect of incorporating defense-in-depth
into design, construction, maintenance and operation is that the facility or system in question
tends to be more tolerant of failures and external challenges.” Therefore, application of
defense-in-depth could mean in part that there is more than one source of cooling water or that
pump make-up can be provided by both electric as well as direct-drive diesel pumps.
Additionally, defense-in-depth can mean that even if a serious outcome (such as fuel damage)
occurs, there is further protection, such as containment for operating plants to prevent
radionuclide releases to the environment and emergency response measures to provide dose
savings to the public.

The defense-in-depth philosophy applies to the operation of the SFP in a decommissioning
plant. The philosophy also applies to the potential regulatory changes contemplated for
decommissioning plants. Implementation of defense-in-depth for SFPs is different than for
nuclear reactors because the hazards are different. The robust structural design of a fuel pool,
coupled with the simple nature of the pool support systems, goes far toward preventing
accidents associated with loss of water inventory or pool heat removal. Additionally, because
the essentially quiescent (low-temperature, low-pressure) initial state of the SFP and the long
time available for taking corrective action associated with most release scenarios provide
significant safety margin, a containment structure is not considered necessary as an additional
barrier to provide an adequate level of protection to the public. Likewise, the slow evolution of

4-6


Enhanced Seismic Checklist
Item 1:
Requirement: Identify Preexisting Concrete and Liner Plate Degradation

Basis: A detailed review of plant records concerning spent fuel pool concrete and liner plate
degradation should be performed and supplemented by a detailed walkdown of the accessible
portions of the spent fuel pool concrete and liner plate. The purpose of the records review and
visual inspection activities is to accurately assess the material condition of the SFP concrete

- and liner in order to assure that these existing material conditions are properly factored into the
remaining seismic screening assessments.

Design Feature: The material condition of the SFP concrete and liner, based upon the
records review and the walkdown inspection, will be documented and used as an engineering
input to the following seismic screening assessments.

Item 2:
Requirement: Assure Adequate Ductility of Shear Wall Structures

Basis: The expert panel involved with the development of Reference 1 concluded that, “ For the
Category 1 structures which comply with the requirements of either ACI 318-71 or ACI 349-76 or
later building codes and are designed for an SSE of at least 0.1g pga, as long as they do not
have any special problems as discussed below, the HCLPF capacity is at least 0.5g pga.” This
conclusion was based upon the assumption that the shear wall structure will respond in a ductile
manner. The “special problems” cited deal with individual plant details which could prevent a
particular plant from responding in the required ductile fashion. Examples cited in Reference 1
included an embedded structural steel frame in a common shear wail at the Zion plant (which
was assumed to fail in brittle manner due to a potential shear failure of the attached shear studs)
and large openings in a “crib house” roof (also at the Zion plant) which could interrupt the
continuity of the structural slab.

Other examples which could impact the ductility of the spent fuel pool structure include
large openings which are not adequately reinforced or reinforcing bars that are not sufficiently
embedded to prevent a bond failure before the yield capacity of the steel is reached.

Design Feature: This design feature requirement will be documented based on a review of
drawings and a SFP walkdown.

Item 3:

Requirement: Assure Design adequacy of Diaphragms (including roofs)

Basis: In the design of many nuclear power plants, the seismic design of roof and floor
diaphragms has often not received the same level of attention as have the shear walls of
the structures. Major cutouts for hatches or for pipe and electrical chases may pose

special problems for diaphragms. Since more equipment tends to be anchored to the
diaphragm compared to shear walls, moderate amounts of damage may be more critical

A2B-7


Table 3.2 Spent Fuel Pool Cooling Risk Analysis — Frequency Partition for Air Flow

% LOW AIR FLOW
SEQUENCES % HIGH AIR FLOW {ADIABATIC)
Seismic 30% 70%
Heavy load drop 50% 50%
Loss of offsite power, severe 90% 10%
weather

In Table 3.2 for seismic sequences, we have assumed that 30 percent of the time the building
will turn over two building volumes of air per hour (high air flow) and 70 percent of the time the
individual fuel bundle of concern will receive little or no air cooling. These percentages are
based on discussions with staff structural engineers, who believe that, at accelerations in excess
of 1.2 g spectral acceleration (which is greater than three times the safe shutdown earthquake
(SSE) for many reactor sites east of the Rocky Mountains), there is a high likelihood of building
damage that blocks air flow. For heavy load drop sequences, the staff assumed a 50 percent
partition for the high air flow case. This is based on considering both damage to fuel bundles
due to a heavy load drop that renders bundles uncoolable and the alternative possibility that the
drop damages the building structure in a way that blocks some spent fuel bundles. For loss of
offsite power events caused by severe weather, the staff assumed a 90 percent partition for the
high airflow case. This is based on a staff assumption that openings in the SFP building (e.g.,
doors and roof hatches) are large enough that, if forced circulation is lost, natural circulation
cooling will provide at least two building volume of air per hour to the SFP. This assumption
may need to be confirmed on a plant-specific basis. The staff did not partition the rest of the
sequences in Table 3.2, since their absolute value and contribution to the overall zirconium fire
frequency are so low.

The frequency partitioning shows that a large portion of the seismic and heavy load drop
sequences have low air flow. This partitioning did not consider the possibility that the air flow
path is blocked by residual water in the SFP. When the potential for flow blockage by residual
water is considered, an even greater portion of the events would result in no (or low) air flow.
Fuel heatup calculations in Section 2 show that for the first several years after shutdown, the
fuel heatup time (e.g., time to reach 900 °C) for the adiabatic and air-cooled cases is
comparable. Thus, the effects of partitioning are negligible for this period. Because SFP and
SFP building fragilities and failure modes are plant-specific, and the heatup time for the
adiabatic and air-cooled cases differ only slightly, the staff did not consider the partitioning in
estimating the frequency of SFP fires. Whether or not a spent fuel bundle receives high air flow
or low air flow fuel uncovery does not change our insights into the risk associated with operation
of SFPs.

3-10


UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS
PENALTY FOR PRIVATE USE, $300


the probability of failure to repair the pump would be Exp [-(1/10) ( 72] = 1.0E-3. This event is
modeled in the fault tree, LP1-OMK-L.

HEP-FW-REP-DEPEN represents the failure of the repair crew to repair a firewater pump. Note
that repair was not credited for top event OCS; however, it has been assumed that the repair
crew would have made an attempt to restore the SFPC system, and so dependency was
modeled in the failure to repair the firewater system. A probability of failure to repair a pump in
88 hrs is estimated to be 1.0E-3. For HEP-FW-REP-DEPEN a low level of dependence was
applied modifying the failure rate of 1.0E-3 to 5.0E-2 using the THERP formulation for low
dependence. This event is modeled in the fault tree, LP1-OMK-U.

In addition, in fault tree LP1-OMK-U, the possibility that no action is taken has been included by
incorporating an AND gate with basic events HEP-DIAG-SFPLPI and HEP-RECG-DEPEN. The
latter is quantified on the assumption of a low dependency.

Hardware Failure Probabilities

In the case of LP1-OMK-U, both firewater pumps are available. Failure of both firewater pumps
is represented by basic event FP-2PUMPS-FTF. In the case of LP1-OMK-L, only the
diesel-driven firewater pump is available, and its failure is represented by basic event
FP-DGPUMP-FTF.

The pump may be required to run 8 to 10 hours at the most (250 gpm capacity), given that the
water inventory drops by 20 ft (i.e., 3 ft above the top of the fuel). A failure probability of 3.7E-3
for failure to start and run for the electric pump and 0.18 for the diesel driven pump are used
from INEL-96/0334. These individual pump failures result in a value of 0.18 for event
FP-DGPUMP-FTF and 6.7E-4 for event FP-2PUMPS-FTF.

4.3.4.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-RECG-DEPEN 5.0E-02
HEP-RECG-FWSTART 2.0E-5
HEP-FW-START 4.0E-5
HEP-FW-REP-DEPEN 5.0E-2
HEP-FW-REP-NODEP 1.0E-3
FP-2PUMPS-FTF 6.7E-4
FP-DGPUMP-FTF 1.8E-1

Appendix 2A A2A-39


estimated value range was 1.0x10* to 1.0x10"° per year. The average value was estimated to
be 7.0x10* per year. This value does not credit onsite or offsite recovery actions.

As a check, we calculated the frequency for loss of a support system supply based on the DOE
model, including wing and skid area, for a 10 x 10 x 10-foot structure. The estimated frequency
range was 1.1x10° to 1.1x10° per year with the wing and skid area modeled, with the average
estimated to be 7.3x10” per year. Using the point model, the estimated value range was
2.4x10°? to 1.1x10° per year, with the average estimated to be 7.4x10"° per year. This value
does not credit onsite or offsite recovery actions.

5. UNCERTAINTIES

Mark-! and Mark-Il secondary containments do not appear to have any significant structures that
would reduce the likelihood of penetration, although on one side there may be a reduced
likelihood because of other structures. Mark-lll secondary containments may reduce the
likelihood of penetration, since the SFP may be considered to be protected by additional
structures.

6. REFERENCES

1. DOE-STD-3014-96, "Accident Analysis for Aircraft Crash Into Hazardous Facilities," U.S.
Department of Energy (DOE), October 1996

2. A. Mosleh and R.A. Bari (eds),"Probabilistic Safety Assessment and Management,”
Proceedings of the 4th International Conference on Probabilistic Safety Assessment and
Management, PSAM 4, Volume 3, 13-18 September 1998, New York City.

3. C.T. Kimura et al.,"Aircraft Crash Hit Analysis of the Decontamination and Waste
Treatment Facility (DWTF), Lawrence Livermore National Laboratory.

4. K. Jamali, et al., "Application of Aircraft Crash Hazard Assessment Methods to Various
Facilities in the Nuclear Industry."

5. NUREG/CR-5042, "Evaluation of External Hazards to Nuclear Power Plants in the
United States," Lawrence Livermore National Laboratory, December 1987.

Appendix 2D A2D-4


1.0 INTRODUCTION

Decommissioning plants have requested exemptions to certain regulations as a result of their
permanently defueled condition. Although the current Part 50 regulatory requirements
(developed for operating reactors) ensure safety at the decommissioning facility, some of these
requirements may be excessive and not substantially contribute to public safety. Areas where
regulatory relief has been requested in the past include exemptions from offsite emergency
planning (EP), insurance, and safeguards requirements. Requests for consideration of changes
in regulatory requirements are appropriate since the traditional accident sequences that
dominate operating reactor risk are no longer applicable. For a defueled reactor in
decommissioning status, public risk is predominantly from potential accidents involving spent
fuel. Spent fuel can be stored in the spent fuel poo! (SFP) for considerable periods of time, as
remaining portions of the plant continue through decommissioning and disassembly. To date,
exemptions have been requested and granted on a plant-specific basis. This has resulted in
some inconsistency in the scope of evaluations and the acceptance criteria applied in
processing the exemption requests.

To improve regulatory consistency and predictability, the NRC undertook this effort to improve
the regulatory framework applicable to decommissioning plants. This framework utilized risk-
informed approaches to identify the design and operational features necessary to ensure that
risks to the public from these shutdown facilities are sufficiently small. This framework forms a
technical foundation to be used as one input to developing regulatory changes, as well as a part
of the basis for requesting and approving exemption requests until rulemaking is completed.

In support of this objective, the NRC staff has completed an assessment of SFP risks. This
assessment utilized probabilistic risk assessment (PRA) methods and was developed from
analytical studies in the areas of thermal hydraulics, reactivity, systems analysis, human
reliability analysis, seismic and structural analysis, external hazards assessment, and offsite
radiological consequences. The focus of the risk assessment was to identify potential severe
accident scenarios at decommissioning plants and to estimate the likelihood and consequences
of these scenarios. The staff also examined the offsite EP for decommissioning plants using an
analysis strategy consistent with the principles of Regulatory Guide (RG) 1.174.

Preliminary versions of this study were issued for public comment and technical review in June
1999 and February 2000. Comments received from stakeholders and other technical reviewers
have been considered in preparing this assessment. Quality assessment of the staffs
preliminary analysis has been aided by a small panel of human reliability analysis (HRA) experts
who evaluated the human performance analysis assumptions, methods and modeling. A broad
quality review was carried out at the Idaho National Engineering and Environmental Laboratory
(INEEL).

The study provides insights for the design and operation of SFP cooling and inventory makeup
systems and practices and procedures necessary to ensure high levels of operator performance
during off-normal conditions. The study concludes that, with the fulfillment of industry
commitments and satisfaction of a number of important staff assumptions, the risks from SFPs
can be sufficiently low to evaluate exemptions involving small changes to risk parameters and to
contribute to the basis for related rulemaking.

1-1


calculated using THERP. The relevant performance shaping factors included greater than 24
hours for diagnosis, high stress, well-trained operators, diagnostic procedures, and good work
processes. A low dependence for the walk-down error was applied.

Because it is assumed that at most 9 hours are available, no credit was given for repair of the
SFP cooling system.

Non-HEP Probabilities

Fault tree LP1-OCS represents failure of the SFP cooling system to restart and run. Hardware
failure rates have been taken from INEL-96/0334 (Ref. 15). It is assumed that SFPC system will
be maintained since it is required to be running all the time.

4.3.3.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-DIAG-SFPLP1 1.0E-06
HEP-SFP-STR-LP1 5.0E-6
SPC-CKV-CCF-H 1.9E-5
SPC-CKV-CCF-M 3.2E-5
SPC-HTX-CCF 1.9E-5
SPC-HTX-FTR 2.4E-4
SPC-HTX-PLG 2.2E-5
SPC-PMP-CCF 5.9E-4
SPC-PMP-FTF-1 3.9E-3
SPC-PMP-FTF-2 3.9E-3

4.3.4 Top Event OMK - Operator Recovery Using Makeup Systems
4.3.4.1 Event Description and Timing

This top event represents the failure to provide make-up using the firewater pumps. If offsite
power is recovered then the fault tree LP1-OMK-U represents this top event. In this case, the
operator has both electric and diesel firewater pumps available. If offsite power is not recovered
then fault tree LP1-OMK-L represents this top event. In this case, the operator has only the
diesel firewater pump available.

4.3.4.2 Relevant Assumptions

. It is assumed that the procedures guide the operators to wait until it is clear that SFP
cooling cannot be reestablished (e.g., using cues such as the level drops to below the
suction of the cooling system or the pool begins boiling) before using alternate makeup
sources. Therefore, they have 88 hours to start a firewater pump.

e There is a means to remotely align a makeup source to the SFP without entry to the refuel

Appendix 2A A2A-37


Response: When the NRC issued decommissioning regulations in 1988, it was assumed that
decommissioning would normally take place after the facility's operating license expired. The
licensee was obligated to submit a preliminary decommissioning pian 5 years before the
license expired. The preliminary decommissioning plan contained a cost estimate for
decommissioning and an up-to-date technical assessment of the factors that could affect
planning for decommissioning. This included (1) the decommissioning alternative selected, (2)
the major technical actions necessary to carry out decommissioning safely, (3) the current
situation with regard to disposal of high-level and low-level radioactive waste, (4) the residual
radioactivity criteria, and (5) other site-specific factors that could affect decommissioning
planning and cost.

The 1988 rule also required that no later than 1 year before expiration of the license (or within 2
years of permanent cessation of operations for plants closing before their license expires), a
licensee had to submit an application for authority to decommission the facility. The application
was to be accompanied by or preceded by a proposed decommissioning plan. The proposed
decommissioning plan was to include (1) the alternative selected for decommissioning with a
description of the activities involved, (2) a description of controls and limits on procedures and
equipment to protect occupational and public health and safety, (3) a description of the planned
final radiation survey, (4) an updated cost estimate for the chosen alternative and a pian for
ensuring the availability of adequate funding, and (5) a description of the technical
specifications, quality assurance provisions, and physical security plan provisions in place
during decommissioning. A supplemental environmental report that described any substantive
environmental impacts that were anticipated but not already covered in other environmental
impact documents was also required.

The NRC would review the decommissioning plan and would approve it by issuing an order if
the plan demonstrated that the decommissioning would be performed in accordance with
regulations and there were no security, health, or safety issues. The NRC would also require
that notice be given to interested persons. However, the NRC could add other conditions and
limits to the plan that it deemed appropriate. The license would then be terminated if the NRC
determined that the decommissioning had been performed in accordance with the approved
decommissioning plan and the order authorizing decommissioning, and if the final radiation
survey and associated documentation demonstrated that the facility and site were suitable for
release for unrestricted use.

In August 1996 the regulations were revised for several reasons. First, the experience gained
in the early decommissioning activities associated with several facilities did not reveal any
activities that required NRC review and approval of a decommissioning plan. Second,
environmental impacts associated with decommissioning those early facilities resulted in
impacts consistent with those evaluated in the "Generic Environmental impact Statement on
Decommissioning of Nuclear Facilities," NUREG-0586. And finally, experience gained from
reviewing numerous decommissioning oversight activities at a number of these facilities also
indicated that the decommissioning activities were in general no more complicated than
activities normally undertaken at operating reactors without prior and specific NRC approval.
The revised rule redefined the decommissioning process and required licensees to provide the
NRC with early notification of planned decommissioning activities at their facilities went into
effect. The rule made the decommissioning process more efficient and uniform. It provided for
greater public awareness and clarified the opportunity for participation in the decommissioning
process. It also gave plant personnel a clearer understanding of the process for changing from
an operating organization to a decommissioning organization.

Appendix 6 A6-33


Comment #36: Were heavy objects, such as crane rail or masonry wall, falling into the SFP or
taking out electricity during decommissioning activities addressed in the study?

Response: The loss of electricity and the control of heavy loads were considered in the study.
The loss of electricity would result in a loss of the spent fuel pool cooling system. IDC #1 and
SDA #5 deal with controlling heavy loads over the spent fuel pool. SDA #6 requires that
licensees complete a seismic checklist that evaluates that seismic robustness of the pool and
building structure. If a plant cannot successfully complete the generic seismic checklist, then a
site-specific assessment would be performed.

Comment #37: Since the National Severe Storm Center is predicting more frequent and more
intense severe weather phenomena, shouldn't the size and velocity of wind-driven missiles and
maximum height of storm surges be reassessed?

Response: The licensing basis for severe weather phenomena is conservative, such that
increases in storm intensity should still be bounded by the current licensing basis. When
severe weather is expected, such as hurricanes, the NRC monitors the licensee’s readiness
and performance closely. If more severe storms were occurring or a plant (or plants) did not
function as expected, we would evaluate the need to update plants’ storm-related analyses and
communicate with industry, such as using information notices and bulletins, to ensure that
licensees were ‘aware of events that occur at other plants and our expectations of their
performance. Also, if a licensee requests a change to its licensing basis dealing with storms,
such as tornados, or storm-generated missiles, then they would look at more recent data
collected since the licensing of the plant.

Comments #38 and 39: (A) All pools leak, dry storage is the only way for long term safety.
(B) The NRC should identify all SFP’s that leak. Degradation of the liner and concrete should
be investigated. The leaks should be sealed.

Response: The staff has determined that both wet storage (in a SFP) and dry storage (in
casks) are safe methods to hold spent fuel. Most pools have a leak detection system between
the steel liner and the concrete wall to identify and quantify if leakage from the liner occurs.
This is not leakage to the environment. This water is collected by the system in the plant. This
system allows licensees to monitor a situation and evaluate if there is a safety concern. Two
plants do have leaking spent fuel pools. The licensees are closely monitoring the leak to
ensure that there is no public hazard. Dry storage casks are a viable option for spent fuel
storage for licensees. Dry storage casks are currently approved for fuel that have been
removed from the reactor for at least five years. Many licensees are choosing to use dry cask
storage in addition to the spent fuel pool.

Comment #40: What happened to the commitment verbally agreed upon through a public
stakeholder to install a single failure proof crane system using safety grade electrical
equipment?

Response: The staff reviewed the transcript of the public meeting. NEI verbally committed
decommissioning plants to implement Phase II of NUREG-0612 (Control of Heavy Loads),
which prescribed the use of single failure proof cranes or to implement a load drop analysis.
NEI provided this commitment in writing on November 12, 1999 (See Appendix 5). The
commitment was included in the analysis and documented in the report as IDC #1.

Appendix 6 A6-12


that assures that the good practices found at decommissioning sites visited by the staff will be
implemented at future decommissioning sites. Some industry commitments and staff
assumptions, such as IDC #1 and SDAs #2, #3, and #4, may enhance the capabilities currently
practiced by existing decommissioning plants. Where possible (e.g., for some initiating event
frequencies), the staff has used actual data from spent fuel pool events. The industry
commitments and staff assumptions provide a basis for the staff's conclusion that the low
human error probabilities associated with the loss of SFP cooling and loss of inventory events
are justified. In addition, the commitments provide a bound on the risk associated with the two
events that could rapidly drain the spent fuel pool (i.e., seismic and heavy load drop events.)

Comment #33: The staff noted a recent event (January 2000) that occurred during shutdown,
when SFP monitoring should have been a priority. This event should have raised the initiating
event frequencies, not lowered them.

Response: The staff agrees that including the two recent loss-of-cooling events mentioned the
draft report would increase the initiating event frequency for loss of cooling accidents.
However, since the fuel uncovery frequency from this event is very low (approximately 10° per
year), the conclusion in the report that the loss of cooling events are not a major risk
contributors is not affected. However, these recent events illustrate the importance of industry
commitments, particularly IDC #5, which requires temperature instrumentation and alarms in
the control room. Additionally, SDA #3 calls for direct measurement of water temperature and
level in the spent fuel pool.

Comment #34: The discussion in Section 3.3.2 [of the draft report] states that many of the
events listed in NUREG-1275, Volume 12, do not apply to a decommissioning facility.
Therefore, adherence to IDCs #2, 5, 8, and 10 are not really important to establishing a low
frequency of fuel uncovery. .

Response: The commenter correctly noted that many of the initiating events from operating
reactor spent fuel pool incidents that are discussed in NUREG-1275 do not apply to
decommissioning facilities. The staff likewise did not include these events when estimating the
frequency of events at decommissioning plants. To help assure that the frequency of these
events does not end up being much higher than assumed by the staff in its risk assessment,
the industry committed to IDCs and SDAs including those mentioned regarding procedures and
planning for contingencies to limit, prevent, or mitigate loss of inventory and loss of cooling
events.

Comment #35: How did the staff come up with the factor of 100 reduction in the failure rate for
heavy load drops for single-failure-proof systems?

Response: For a non-single-failure-proof handling system, the mean probability of a loss-of-
inventory event was estimated based on NUREG-0612. In NUREG-0612, an alternate fault
tree (Figure B-2, page B-16) was used to estimate the probability of exceeding the release
guidelines (loss-of-inventory) for a non-single failure proof system. The mean value was
estimated to be about 2.1x10° per year when corrected for the new Navy data and 100 lifts per
year. A comparison of this mean value to the 2.0x107 per year mean value for the single-
failure-proof crane shows a factor of 100 reduction.

Appendix 6 AG-11


Response: Foreign objects that would fall between or into the fuel assemblies to be close to
the fuel would be of a small size and would have an insignificant effect on the heat input in
comparison to the spent fuel. As part of this study, the staff performed a thermal hydraulic
sensitivity study assuming adiabatic conditions (no heat removal), which would produce a
faster heatup of the fuel than would actually be expected. Additionally, licensees have
programs to keep foreign objects from entering the spent fuel pool. Retrievable foreign objects
that fall into the pool are moved to designated storage areas within the pool.

Comment #70: The energy of reaction for air oxidation in the draft report is incorrect.
Response: The staff confirmed that the draft report is correct. It appears that the commenter

based the calculation on 92 kg of zirconium in a mole, when there are 92 grams of zirconium in
amole.

Appendix 6 A6-23


FAILURE TO RECOVER
‘OFFSITE SOURCES

FAILS TO RECOGNIZE LOSS OF SFPC SYSTEM FAILURE TO RECOVER
NEED TO START FW RECOGNITION FALURE MIAGEFSITE SOURCES
He. TART wpropus HEPANTOPFSTE
rc 1
OPERATORS FAIL TO
RECOGNZENEED TO pea sseslig
COOL og ave GIVEN LOSP EVENT
HERREOSDEPEN HEPDIASSFPLPA
1
FALURETOREOOER
IEE
ae
ae RERRRS
QD. Q.



CONSEQUENCES

Comment #8: Is a gap release considered to give moderate off-site consequences at the time
when a zirconium fire is no longer a threat?

Response: As time increases since permanent shutdown, the fission product inventory
available for release gets smaller and the decay power decreases. As a result, there may not
be sufficient energy to carry released fission products out of the spent fuel pool and offsite.
NUREG/CR-4982, Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue
82, July 1987, provides societal doses for SFP accidents involving a fuel melt release and a
gap release at one year after final shutdown. These societal doses, which are for the
population within 50 miles, are 3x10* rem and 4 rem for a fuel melt release and a gap release.
A gap release is expected to give negligible off-site radiological consequences at this time.
This study did not calculate the consequences of a gap release.

Comment #9: The draft study does not explain the regulatory basis for using 4 rem over 5
years as the threshold dose for relocation.

Response: The Environmental Protection Agency (EPA) report 400-R-92-001, Manual of
Protective Action Guides and Protective Actions for Nuclear Incidents, May 1992, states that,
after the early phase of a nuclear incident, protective actions should be taken to limit the dose
received by an individual to 2 rem in the first year, 0.5 rem/year after the first year, and 5 rem
over 50 years. These Protective Action Guides are implemented in the MACCS code for
relocation criteria by limiting the dose to 4 rem over 5 years, that is, 2 rem in the first year plus
0.5 rem for each of the second through fifth years.

Appendix 6 A6-3


Even with these industry commitments, the staff had to assume the drop in pool inventory due to
loss of inventory events is limited in order to arrive at the low accident frequency calculated for
this scenario. This additional assumption is identified by the staff as SDA #4.

SDA #4 The licensee has determined that the SFP has no drain paths that could lower the
pool level (by draining, suction, or pumping) more than 15 feet below the normal pool
operating level and that the licensee initiates recovery using offsite sources.

3.4.3. Loss of Offsite Power from Plant-Centered and Grid Related Events

A loss of offsite power from plant-centered events typically involves hardware failures, design
deficiencies, human errors (in maintenance and switching), localized weather-induced faults
(e.g., lightning), or combinations. Grid-related offsite power events are caused by problems in
the offsite power grid. With the loss of offsite power (onsite power is lost too, since the staff
assumes no diesel generator is available to pick up the necessary electrical loads), there is no
effective way of removing heat fromthe SFP. If power is not restored in time, the pool will
heatup and boiloff inventory until the fuel is uncovered. The diesel-driven fire pump is available
to provide inventory makeup. If the diesel-driven pump fails, and if offsite power is not
recovered promply, recovery using offsite fire engines is a possibility. Recovery times are the
same as for loss of cooling (discussed in Section 3.4.1).

Even after recovering offsite power, the fuel handlers have to restart the fuel pool cooling
pumps. Failure to do this or failure of the equipment to restart will necessitate other fuel handler
recovery actions. Again, considerable time is available.

The calculated fuel uncovery frequency for this sequence of events is 2.9x10° per year. This
frequency is very low and, as with loss of pool cooling and loss of inventory, is based on
adherence to IDCs #2, #5, #8, and #10. In addition, regular plant walkdowns, clear and explicit
procedures, fuel handler training (SDA #2), and the direct measurement of level and
temperature in the SFP (SDA #3) are assumed as documented.

3.4.4 Loss of Offsite Power from Severe Weather Events

This event represents the loss of SFP cooling because of a loss of offsite power from severe
weather-related events (hurricanes, snow and wind, ice, wind and salt, wind, and one tornado
event). Because of the potential for severe localized damage, tornadoes are analyzed
separately in Appendix 2E. The analysis is summarized in Section 3.5.3 of this study.

Until offsite power is recovered, the electrical pumps are unavailable and the diesel-driven fire
pump is available only for makeup. Recovery of offsite power after severe weather events is
assumed to be less probable than after grid-related and plant-centered events. In addition, it is
more difficult for offsite help to reach the site.

The calculated fuel uncovery frequency for this event is 1.1x10” per year. As in the previous
cases, this estimate was based on IDCs #2, #5, #8, #10 and on assumptions documented in
SDA #2 and SDA #3. In addition, IDC #3, the commitment to have procedures in place for
communications between onsite and offsite organizations during severe weather, is also

3-15


For HEP-FW-START-LOI it was assumed that the operators are under high stress, are engaged
in a highly complex task because of its non-routine nature, have a high level of training, have a
diagnostic procedure, and are a part of a crew that interacts well in a quality facility.

Basic event HEP-FW-REP-NODS (see fault tree, OIL-OMK-SMLKL) represents the failure of the
repair crew to repair a firewater pump for the small leak scenarios. Note that repairing the SFP
regular makeup system is not modeled, as there would not be enough time to get help before
the SFP make-up would be ineffectual and therefore no dependency was modeled in the failure
to repair the firewater system. It is assumed that the operators will focus their recovery efforts on
only one pump. Assuming that it takes another 16 hours before technical help and parts arrive,
the operators have about 49 hours (65 hours less 16 hours) to repair the pump. Therefore,
assuming a 10-hour mean time to repair, the probability of failure to repair the pump would be
Exp (-(1/10) * 49) = 7.5E-3 in the case of a small break scenario.

Basic event HEP-FW-REP-NODLG represents the failure of the repair crew to repair a firewater
pump for the large leak scenarios. For this case there would only be 24 hours to repair the
pump. Therefore, assuming a 10-hour mean time to repair, the probability of failure to repair the
pump would be Exp (-(1/10) * 24) = 9.0E-2 in the case of a large break scenario.

Hardware Failure Probabilities

Failure of both firewater pumps is represented by basic event FP-2PUMPS-FTF. The pump
may be required to run 8 to 10 hours at the most (250 gpm capacity), given that the water
inventory drops by 20 ft (i.e., 3 ft from the top of the fuel). A failure probability of 3.7E-3 for
failure to start and run for the electric pump and 0.18 for the diesel driven pump are used from
INEL-96/0334. These individual pump failures result in a value 6.7E-4 for basic event
FP-2PUMPS-FTF.

4.5.7.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-RECG-FWSTART. 2.0E-5
HEP-RECG-FWSTART-L 5.0E-02
HEP-FW-START 1.0E-5
HEP-FW-REP-NODSM . 7.5E-3
HEP-FW-REP-NODLG 9.0E-2
FP-2PUMPS-FTF 6.7E-4
HEP-RECG-FW-LO! 2.0E-4
HEP-FW-START-LOI 1.3E-3

Appendix 2A A2A-62


Table A4-1 Radionuclide Inventories in the Millstone 1 Spent Fuel Pool

Spent Fuel Pool Inventory (Ci)
30 days after { 90 days after | 1 year after
last last last

Radionuclide | Half-Life | discharge discharge discharge

Co-58 70.9d 2.29E4 1.26E4 8.54E2
Co-60 5.3y 3.72E5 3.15E5 2.85E5
Kr-85 10.8y 1.41E6 1.39E6 1.33E6
Rb-86 18.7d 1.01E4 1.05E3 3.84E-2
Sr-89 50.5d 8.39E6 3.63E6 8.33E4
Sr-90 28.8y 1.42E7 1.42E7 1.39E7
Y-90 28.8y 1.43E7 1.42E7 1.39E7
Y-91 58.5d 1.18E7 5.75E6 2.21E5
2r-95 64.0d 1.94E7 1.00E7 5.10E5
Nb-95 64.0d 2.54E7 1.70E7 1.11E6
Mo-99 2.7d 1.49E4 3.12E-3 0
Tc-99m 2.70 1.43E4 3.01E-3 0
Ru-103 37.3d 1.53E7 5.21E6 4.07E4
Ru-106 1.0y 1.72E7 1.53E7 9.13E6
Sb-127 3.8d 8.21E3 1.39E-1 0
Te-127 109d 2.21E5 1.45E5 2.52E4
Te-127m 109d 2.18E5 1.48E5 2.57E4
Te-129 33.6d 2.74E5 7.79E4 2.68E2
Te-129m 33.6d 4.21E5 1.20E5 4.12E2
Te-132 3.2d 3.7464 8.64E-2 ie)
1-131 8.0d 1.22E6 6.35E3 0
1-132 3.2d 3.85E4 8.90E-2 9)
Xe-133 5.2d 7.29E5 2.30E2 0
Cs-134 2.ty 7.90E6 7.47E6 5.80E6
Cs-136 13.2d 2.05E5 8.13E3 3.91E-3

Appendix 4 A4-3


FAILURE TO MAKEUP
INVENTORY USINS
EIDGFIRE PUMPS

=r 1
LOSS OF SFFCSYSTEM FAILS TODAGNOSE FALURE TO MAKEUP GPEATORFALS
BY TO ESTABLISH
RECOGNITION FALUFES, NEED TO START BW vse dE pote TON oa
ulus HEPRECEPWST-SW amides HERA STARTSW
OPERATOR FAILSTO FAILURE TORECOGNZE (CLESELAND ELECTRIC PASTORS AREY
SNZE NEED" RE PROTECTION PUMPS BLECTRICOR OE SEL
‘GIVEN LOSP(SW) GIVEN PRIOR FALURE FAIL TOSTART ANDRUN- PUMP DETERDENCY
HERCASSrPLr2 HEP-FEOGOEPEN FRPUMPSETE HERAWREPOPSW
1
FAIORE TOMAKELP
USING
OGFIREPUMP
LPZOMeL,
r f 1
CPEATORFALS PROV
TOESTABUSH FAILSTO DIAGNOSE FALSTO PROVIDE
NEED TOSTART FW MAKEUP USING DES.
Qansveney FIRE PROTECTION PUMP
HEP-PWSTART-SW HERREOG FWST-SW
DIESEL FIRE PROTECTION FALS TOREPAR
PUMPFAILSTO STRAT, DIESEL PUMP
ND RUN NODEPENDENCY
(HRS-SW)
FPOGAMRFTF HERAW-RERNODSW

A-13



THERMAL HYDRAULICS

Comment #64: The draft study is deficient in that it ignores the phenomenon associated with
partial draindown of SFP that will suppress convective heat transfer by presence of residual
water at the base of fuel assemblies.

Response: The staff agrees that the partial drain down scenario should be considered and has
included the scenario in the thermal hydraulic analyses in this study.

Comment #65: The draft study is deficient in that partial draindown will lead to a steam-
zirconium reaction producing hydrogen gas which could reach explosive concentrations in the
atmosphere of the spent fuel building, potentially leading to a breach of that building.

Response: The staff agrees that, in a partial draindown scenario, hydrogen would be
produced. The hydrogen concentrations or the consequences of any subsequent hydrogen
burn or explosion have not been calculated for this study. However, the staff believes that the
consequences are bounded by the zirconium fire consequences because no credit is given in
the staff's analysis for building integrity to retain fission products. In effect, a complete building
breach is assumed in the staff's consequence analysis.

Comment #66: Depending on fuel burnup/storage array details, the development of standard
methods is needed for consistent application of regulations.

Response: The staff agrees that a standard methodology or guidelines for thermal hydraulic
analysis would assist in the consistent application of regulations. After a rulemaking plan is
approved by the Commission, the staff will evaluate the best methods to develop the necessary
guidance.

Comment #67: The gap release temperature is too conservative for a success criterion.

Response: The gap release temperature is the temperature at which the cladding can blister
and allow gases trapped between the fuel pellets and the cladding to escape. The temperature
criterion for gap release may also be the threshold for releasing fuel fines and ruthenium.
These considerations and others were included in determining the temperature criterion for the
thermal hydraulic analysis. This is discussed in Appendix 1b.

Comment #68: Fire propagation to low powered fuel is unlikely.

Response: As the fuel decays, the involvement in a fire becomes less likely. However,
sufficient research has not been performed to define clear limits of propagation. The staff
therefore assumed the involvement of 3.5 cores in its analysis based upon previous analyses
(NUREG/CR-0649 and NUREG/CR-4982). This assumption is discussed in Section 3 and
Appendix 4 of this report.

Comment #69: Could foreign materials with lower ignition temperatures enter a drained SFP
and catch fire, thus raising the temperature of the spent fuel to the point of rapid zirconium
oxidation?

Appendix 6 AG-22


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A3-17

Appendix 3


PWR High Density Storage Rack Eigenvalue Following Compressive/Expansion

Figure 1

Events

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A3-14

Appendix 3


In addition to comments on the selection of an ignition temperature, the staff received comments
related to the effect of intermetallic reactions and eutectic reactions. With respect to
intermetallic reactions, the melting temperature of aluminum, which is a constituent in BORAL
poison plates in some types of spent fuel storage racks, is approximately 640 °C. Molten
aluminum can dissolve stainless steel and zirconium in an exothermic reaction forming
intermetallic compounds. In the SFP configuration, zircaloy cladding will be covered with an
oxide layer and unless significant fresh metal surface is exposed through exfoliation there will be
no opportunity to interact metallic zircaloy with aluminum (which similarly will be oxidized).
Aluminum and steel will form an intermetallic compound at a temperature of 1150 °C, (Ref. 5)
which is above the temperature criterion selected for fuel damage.

Besides intermetallic compounds, eutectic reactions may take place between pairs of various
reactor materials, e.g., Zr-Inconel (937 °C), Zr-steel (937°C), Zr-Ag-In-Cd (1200 °C), Zr-B,C
(1627 °C), steel-B,C (1150 °C), etc. (Ref. 6 ). Consideration of eutectics and intermetallics is
important from the standpoint of heat addition as well as assuring the structural integrity of the
storage racks and maintaining a coolable configuration. Noting the eutectic and intermetallic
reaction temperatures, however it does not appear that formation of these compounds imposes
any additional temperature limit on the degradation of cladding in an air environment.

Since the temperature criterion is also a surrogate of sorts for the subsequent release of fission
products it is useful to consider the temperature threshold versus temperatures at which
cladding may fail and fission products be released. Cladding is likely to fail by ballooning and
burst in the temperature range of 700-850 °C, resulting in the release of fission products and
fuel fines. At burst, clean Zircalloy metal will also be exposed, leading to an increase in
oxidation although the total amount of metal involved will be limited. Creep failure of the
cladding at or above 600 °C is also a possibility. This temperature limit is roughly associated
with the 10 hour creep rupture time (565 °C) which has been used as a regulatory limit. While
failure of the cladding at these lower temperatures will lead to fission product release, such
release is considerably smaller than that assumed for the cases where the temperature criterion
is exceeded and significant fuel heatup and damage occurs. Low temperature cladding failures
might be expected to produce releases similar to those associated with dry cask accident
conditions as represented in Interim Staff Guidance (ISG)-5. This NRC guidance document
prescribes release fractions for failed fuel (2x10“ for cesium and ruthenium and 3x10° for fuel
fines). Use of these release fractions would reduce the estimated offsite consequences
dramatically from the fuel melt cases, early fatalities would be eliminated and latent cancer
fatalities would be reduced by a factor of 100. As the temperature limit is increased from 600 °C
to 900 °C there is some evidence that ruthenium releases would be increased based on ORNL
test data from unclad pellets. Canadian data indicate though, that in the case of clad fuel the
ruthenium release did not commence until virtually all of the cladding had oxidized. By this point
it might be surmised that the fuel configuration would more closely resemble a debris bed than
intact fuel rods. Selection of a temperature criterion for fuel pool damage also depends on the
intended use, i.e., whether it is intended as the criterion for the 10 hour delay before the onset of
fission product release or whether it is being used as a threshold for long-term fission product
release. If the criteria is being used to judge when 10 hours are available for evacuation, then it
may be argued that a higher temperature could be adopted, one associated with the significant
release of fission products, 1200 °C, since the release of fission products at lower temperatures
will likely be small. However, in air it may be that the oxidation rate above 900 °C is sufficient to

Appendix 1B A1B-5


Basic Event Basic Event Probability
REC-OSP-PC 1E-02

4.3.3. Top Event OCS - Cooling System Restart and Run

4.3.3.1 Event Description and Timing

This top event represents restarting the SFP cooling system, given that Offsite power has been

recovered within 24 hours. There are two electrically operated pumps and the operator can

start either one. If the operator starts the pump that was in operation, no valve alignment would

be required. However, if the operator starts the standby pump, some valve alignment may be

required.

Fault tree LP1-OCS has several basic events: an operator action representing the failure to

establish SFP cooling, and several hardware failures of the system. If power is recovered within

24 hours, the operator has 9 hours to start the system before boil-off starts.

4.3.3.2 Relevant Assumptions

. The operators have 9 hours to start the SFP cooling system.

. The SFP has at least one SFP water temperature monitor, with either direct indication or a
trouble light in the control room (there could also be indications or alarms associated with
pump flow and pressure) (NEI commitment no. 5).

° Procedures exist for response to and recovery from a loss of power, and the operators are
trained in their use (NEI commitment no. 2).

4.3.3.3 Quantification

Human Error Probabilities

Event HEP-SFP-STR-LP1 represents operator failure to restart/realign the SFP cooling system
in 9 hours. The operator can restart the previously running pump and may not have to make
any valve alignment. if he decides to restart the standby pump he may have to make some
valve alignment. The response part of the error was quantified using SPAR. The relevant
performance shaping factors for this event included expansive time, high stress because of
previous failures, moderately complex task because of potential valve lineups, highly trained
staff, good ergonomics (well laid out and labeled matching procedures), and good work process.

A diagnosis error HEP-DIAG-SFPLP1, representing failure of the operators to recognize the
loss of SFP cooling was also included. Success would most likely result from recognition that
the electric pumps stop running once power is lost and require restart following recovery of
power. If the operator fails to make an early diagnosis of loss of SFP cooling, then success
couid still be achieved during walk-downs following the loss of offsite power. Alternatively, if
power is restored, the operator will have alarms available as well. Therefore this value consists
of two errors. The diagnosis error was calculated using SPAR, and the walk-down error was

Appendix 2A A2A-36


G. M. Holahan 2

The results of these sensitivity studies are described in Attachment 1, which was written, at
NRR request, to be incorporated into the final technical study as an appendix. The range of
consequences for a beyond-design-basis spent fuel pool accident occurring one year after final
shutdown is shown below for early evacuation. This range reflects the uncertainty in the
ruthenium and fuel fines release fractions. NRR also requested our assistance in responding
to the public comments on the Draft Final Technical Study of Spent Fuel Pool Accident Risk at
Decommissioning Nuclear Power Plants. Our responses to these comments in the areas of
offsite radiological consequences and emergency response are provided in Attachment 2.

End of Consequences within 100 Miles (Surry population
Range density)
Early Societal Dose Cancer Fatalities
Fatalities (rem)
Lower 005 4x10° 2,000
Upper 5 8x10° 7,000

Recently, NRR requested additional consequence calculations using fission product inventories
at 30 and 90 days and two, five, and ten years after final shutdown to provide additional insight
into the effect of reductions in inventory available for release. We are currently performing
these calculations and expect to provide the results shortly.

References: 1. Memorandum from G. Holahan to T. King dated March 26, 1999
2. Memorandum from A. Thadani to S. Collins dated November 12, 1999

Attachments: 1. Effect of Source Term and Plume-Related Parameters on Consequences
2. Response to Public Comments on the Consequence Assessment

cc: _—_T. Collins
R. Barrett
J. Hannon
J. Wermiel


represented by 3 flow channels with each representing 1/3 of a core. The three previous

4/3 core offloads are modeled separately. The other three channels each model a full core for a
PWR and slightly more than a full core fora BWR. Only an average fuel rod is modeled for
each channel. The model does not allow heat transfer between different powered flow
channels.

Figure 1A-1 shows that for the configuration modeled, and for decay times of less than about

2 years for PWRs and 1.5 years for BWRs (assuming burnup of 60 GWD/MTU), it would take
less than 10 hours for a zirconium fire to start or for significant fission product releases to begin
once the fuel was fully uncovered and the fuel was cooled by an air flow of about two building
volumes per hour. The figure also shows that after 4 years, PWR fuel could reach the point of
fission product release in about 24 hours.

Heatup Time to Release (Air Cooling)
Fuel Burnup of 60 GWD/MTU

40
“I T
£30 a
2 1 LZ PWR
oO

20 -_
£ ee BWR
a 4+
3
: . a
= - =

0 : ' ' . .

0.164 1 2 3 4
shutdown time (years)

Figure 1A-1 Heatup time from 30 °C to 900 °C

The calculations also indicate that about 4-5 years decay is needed before air cooling is
sufficient to preclude a zirconium fire.

The staff considered a number of sequences where the spent fuel might be only partially
uncovered, such as in the case of a rapid partial draindown to a level at or below the top of
active fuel with a slow boiloff of water after the draindown. This could occur if a large breach
occurred in the liner at or below the top of active fuel. The staff has reasoned that for partial
draindown or other cases, the lack of air cooling because of flow blockage combined with the
geometry of the fuel racks result in the decay heat in the fuel rods effectively heating up the
spent fuel in a near adiabatic manner.’ For these cases all the heat generated is retained in the
fuel, its cladding, and the SFP rack structures. Knowledge of the specific heat of the cladding

Appendix 1A A1A-4


APPENDIX 4D
CHANGE IN RISK ASSOCIATED WITH EP RELAXATIONS

Regulatory Guide (RG )1.174 provides guidance on the allowable increase in the frequency of
large early release associated with a proposed change to the licensing basis. In accordance
with RG 1.174, if the baseline LERF is less than 1x10° per year, plant changes can be
approved that increase LERF by up to 1x10° per year. Relaxations in emergency planning
(EP) requirements do not impact the frequency of events involving a large early release (i.e.,
SFP fire frequency) but instead could increase the consequences associated with the large
release. Hence, in applying the ALERF concept to plant changes that impact consequences it
is necessary to translate the allowable increase in LERF into an allowable increase in risk.

The risk increase associated with a ALERF of 1x10* per year can be bounded by considering
the consequences for a worst case large early release sequence, in conjunction with the
maximum allowable frequency increase (i.e., 1x10° per year). This approach provides an
upper limit on the increase in risk that might be approved in accordance with RG 1.174 principle
of permitting only small increases in risk. The allowable increase in risk will be plant specific
since the allowable increase in LERF of 1x10* per year applies to all sites irrespective of such
factors as population and meteorology. However, risk-significant differences between sites will
tend to similarly impact both the SFP and reactor accident consequences. Hence, the
comparisons of SFP risks to the allowable risk increases derived for Surry should be generally
applicable to other sites.

The consequences associated with the source term that produced the greatest number of early
fatalities in the NUREG-1150 study for Surry are provided in Table 1 below. The
consequences are reported separately for internal events and seismic events. The risk
measures reported for seismic events are based on the LLNL hazard curve and are about an
order or magnitude more severe than those based on the EPRI hazard curve. The maximum
allowable level of risk increase is the product of the consequences (in this case, the
consequences for the worst seismic event since it is bounding) and the allowable frequency
increase of 1x10° per year. This risk increase is provided in the last column of Table 1.

It should be noted that the Commission’s Quantitative Health Objectives (QHOs) correspond to
an individual early fatality risk of 5x10’ per year and an individual latent cancer fatality risk of
2x10° per year. Thus, the risk increase values inferred from RG 1.174 for individual early
fatality risk (8.7x10* per year) and individual latent cancer fatality risk (6.9x10* per year)
represent about 17 percent and 4 percent of these QHOs, respectively. This margin reflects
the strategy taken in establishing the acceptance guidelines for risk increase in RG 1.174.
Specifically, in RG 1.174 the NRC adopted more restrictive acceptance guidelines than might
be derived directly from the Commission's Safety Goal Policy Statement. This policy was
adopted to account for uncertainties and for the fact that safety issues continue to emerge
regarding design, construction, and operational matters.

Table 2 summarizes the bases for evacuation modeling for each of the major contributors to
SFP fires. The effectiveness of EP was characterized is such a way to maximize the value of
formal EP in the "full EP" case and minimize the value of ad hoc EP in the “relaxed radiological
preplanning” case. As such, the resulting estimates of the risk increase associated with EP
relaxations represent an upper bound on the potential risk increase.

The consequences associated with each of the events leading to SFP fires are provided in

Table 3 for the "full EP” case and "relaxed radiological preplanning" case. The consequences
are based on results of calculations reported in Appendix 4A. In several cases where MACCS2

Appendix 4D A4D-1


for which the computed seismic risk exceeds 0.5x10° based on the mean EPRI89 hazard
estimates.

$. Conclusions

If based on the mean LLNL93 hazard estimates (Ref. 8 and 9) it is acceptable to have up
to a mean 3x10° annual seismic risk of spent fuel pool failure at the screening level, then
Screening Level 2 defined in Section 3 represents a practical screening level. Only 8 of the 69
sites have computed seismic risks greater than 3x10° at this screening level. Screening Level 2 is
set at a peak 5% damped spectral acceleration (PSA) level of 1.2g (equivalent to a PGA level of
0.5g).

Based on the mean EPRI89 hazard estimates (Ref. 11), Screening Level 2 would
generally result in seismic risk of spent fuel pool failure estimates less than 0.5x10° for spent
fuel pools which passed the screening criteria. Only 4 out of 60 sites have computed seismic
risks greater than 0.5x10° at this screening level.

The screening criteria given in Refs. 4 and 7 represent a good start on developing
screening criteria for spent fuel pools at Screening Level 2. However, I have three significant
concems which are discussed in Sections 3.1 through 3.3. In my judgment, a detailed fragility
review of a few spent fuel pools will be necessary in order to address my concerns. These
reviews should concentrate on aboveground spent fuel pools with walls not backed by soil
backfill. I believe these reviews need to be performed before a set of screening criteria can be
finalized at Screening Level 2.

A2B-19


"SERS
bee

FARSTO! FALURE TORECOVER

‘NED TO START FW WAGFPSTE SOURCES

Seen QQ.

1
FAILURE TO RECOVER
OFFSITE SOURCES
(SMK-LATE)
LOL-OFI 1 L
f 1
FAILS TO RECOGNIZE FAILURE TO RECOVER:
NEED TO START FW MA OFFSITE SOURCES
PUMPS

HEP-RECG-FUSTART-L HEPANV-OFFSITE

A-20



Comment #79: For insurance, the obligation for secondary financial protection should end at
such time that a determination can be made that clad surface temperatures greater than 570
°C can not occur in a dry configuration. The calculation of this temperature should be by
approved methodology. However as supported in the technical report, in the absence of any
calculation, the obligation should end after a period which is less than five years. The capacity
required of primary financial protection should be reduced after the period of time determined
as above for secondary financial protection.

Response: The liability insurance requirements of our regulations are meant to ensure that the
public is protected in the event of a low probability, high consequences event. The accident
sequences that could lead to a zirconium fire are low probability but could result in high
consequences in terms of property damage and land contamination. In this study, a spectrum
of event initiators and sequences of events were analyzed to determine if the events were
creditable. The staff will evaluate the need for analytical calculations and regulatory guidance
during the rulemaking process.

Comment #80: The obligation for decommissioning plants to participate in the secondary
financial protection should be reviewed in light of the low public risk posed for SFPs for
decommissioned plants. Industry does not believe that the risk justifies requiring participation.
(The majority of the 3x10* risk of significant off-site consequences comes from an upper bound
determination of the risk posed by seismic events, not on a best estimate of the seismic risk).

If it is determined that participation [in secondary financial protection] will be required during the
short time that decommissioning plants pose a non-zero risk, then the level of participation
should be in proportion to a best estimate of the risk posed relative to the risk posed by
operating plants. If any participation is required, it should be only for the short period that clad
surface temperatures greater than 570°C can occur in a loss of water configuration. The
calculation of this temperature should be by an approved methodology.

The commenter also stated that the capacity required for primary financial protection should be
eliminated for consideration of any potential for accidents with significant off-site
consequences. For other events with off-site consequences, on-site coverage should be
reduced to $25 million dollars (M) for the period when the spent fuel remains in the pool and
off-site coverage should be reduced to $5-10M. When the fuel has been removed off-site or
placed in an off-site ISFSI, on-site coverage should be reduced to $25M while the site still
contains significant sources of radioactive material. On-site coverage could be reduced to zero
when there are no sources exceeding 1000 gallons of fluid. Off-site coverage should be
reduced to $5-10M for plants with fuel off-site or in an on-site ISFSI.

Response: The underlying purpose of 10 CFR 50.54(w) is to provide sufficient property
damage insurance coverage to ensure funding for on-site post-accident recovery stabilization
and decontamination costs in the unlikely event of a nuclear accident. Section 140.11 of Title
10 of the CFR also serves to provide sufficient liability insurance to ensure funding for claims
resulting from a nuclear incident or precautionary evacuation. The property and liability
insurance requirements of our regulations are meant to ensure that the public is protected in
the event of a low probability, high consequence event.

In SECY-93-127, "Financial Protection Required of Licensees of Large Nuclear Power Plants
During Decommissioning," the staff explains that insurance coverage is necessary for reactor
licensees as determined by "reasonably conceivable" accidents. Reasonably conceivable
accidents may exceed design basis accidents but are less severe than remotely possible

Appendix 6 A6-26


As a measure of whether the risks from SFPs at decommissioning plants were sufficiently low to
allow small changes to risk parameters, the concept of a pool performance guideline (PPG) was
presented in the February 2000 study based on the principles of RG 1.174. In the study, the
staff stated that consequences of an SFP fire are sufficiently severe that the RG 1.174 large
early release frequency baseline of 1x10° per reactor year is an appropriate frequency guideline
for a decommissioning plant SFP risk and a useful measure in combination with other factors
such as accident progression timing, for assessing features, systems, and operator performance
for a spent fuel pool in a decommissioning plant. Like the February 2000 study, this study uses
the PPG of 1x10° per reactor year as the baseline frequency for a zirconium fire in the SFP.

The study is divided into three main parts. The first (Section 2) is a summary of the thermal-
hydraulic analysis performed for SFPs at decommissioning plants. The second (Section 3)
discusses how the principles of risk informed regulation are addressed by proposed changes.
The third (Section 4) discusses the implications of the study for decommissioning regulatory
requirements.


k, = 52.67 exp (-17597/T) {kg/m?}*-sec* [rate constant of O, mass produced]

It was predicted that based on an escalation temperature of 1200 °C in steam (observed in
many of the steam tests), the equivalent heat generation rate in air would produce a
temperature escalation at approximately 925 °C. The above equation for air kinetics was
identified in Reference 3 as the best fit for the CODEX AIT test data, i-e., it provided the best
agreement to the temperature transient in the peak position. For steam kinetics, the rate
equation used in MELCOR was selected for calculating the heat generation rate. The prediction
of an escalation temperature in air using this approach seems to conform quite well with the
observed behavior in the transient CODEX tests and lends further credence to the relative effect
of oxidation in air with respect temperature escalation. The assumption of parabolic kinetics is
routine in oxidation calculations and has been shown to provide a good match with a wide
spectrum of experimental data even though, over select temperature ranges, deviations from
that formulation have been observed. At temperatures above 900 °C, the reaction rate in air is
high, regardless of whether parabolic or linear kinetics is assumed at that point and
distinguishing between the rates of escalation is unimportant for our purposes.

3. ACCEPTANCE CRITERIA

In assessing a temperature criterion for escalation of the oxidation process and subsequent
temperature escalation it is necessary to reconsider the intended uses of the criterion: 1) to
evaluate the decay time after which the fuel heatup, in the case of complete fuel uncovery, leads
to reaching that temperature at 10 hours and 2) to evaluate the decay time after which the fuel
heatup, in the case of complete uncovery will never exceed the temperature criterion.

On balance it appears that a reasonable criteria for the threshold of temperature escalation in an
air environment is a value of approximately 900 °C. This value is supported both by limited
experimental data as well as by inference from the more abundant steam testing data. While
certain weight gain data indicate the onset of a break away in the oxidation rate at lower
temperatures after a period of 10 hours, this additional time period then exceeds the time
interval for which the first use of the criterion is intended. With regard to the second use of the
criterion, determination of the point at which severe fuel heatup is precluded, the onset of
breakaway indicated in certain tests indicates that the temperature criterion should be lowered
to 800 °C. Itis important to stress that, in both instances, the temperature criteria should be
used together with a thermal-hydraulic analysis that considers heat generation (i.e., decay heat
and zircaloy reactions) and heat losses. For the second use of the criterion, i.e., establishing a
threshold for precluding escalation, the analysis must demonstrate that heat losses, through
convection, conduction and radiation, are sufficient to stabilize the temperature at the value
selected.

In the case of slow, complete draining of the pool, or partial draining of the pool it is appropriate
to consider use of a higher temperature criterion for escalation, perhaps as high as 1100 to

4200 °C. This would be appropriate if the primary oxidation reaction was with steam. Such a
temperature criterion is relevant for the first intended use of the criterion, determining the point at
which the temperature is not exceeded for 10 hours, however it is not appropriate for use as a
long-term equilibrium temperature since over long intervals at such high temperature, one might
reasonably expect significant fission product releases.

Appendix 1B A1B-4


FAILURE TO RECOVER:
OFFSITE SOURCES
 WEAT
1
I r 1.
FALURE TORECOVER
FAILS TO RECOGNZE LOSS OF SFPC SYSTEM VIAOFFSITE SOURCES.
NEED TO START FW [RECOGNITION FALLURE (SEVERE WEATHER)
Sw HEP-INV- SV
FALS TO Fi TO.
NEED TOCOGL POOL
RECOGNIZE caitg PALURE
HEP.
\
rey
(EERE WEATIER)
r 1
TAME ROOE
ERENT QUmeEas

A-14


APPENDIX 2
ASSESSMENT OF SPENT FUEL POOL RISK AT DECOMMISSIONING PLANTS

1.0 INTRODUCTION

As the number of decommissioning plants increases, the ability to address generic regulatory
issues has become more important. After a nuclear power plant is permanently shut down and
the reactor is defueled, most of the accident sequences that normally dominate operating
reactor risk are no longer applicable. The predominant source of risk remaining at permanently
shutdown plants involves accidents associated with spent fuel stored in the spent fuel pool
(SFP). Previously, requests for relief from regulatory requirements that are less safety
significant for decommissioning plants than operating reactors were granted on a plant-specific
basis. This is not the best use of resources and led to differing requirements among
decommissioning plants. The NRC Commission urged its staff to develop a risk-informed basis
for making decisions on exemption requests and to develop a technical basis for rulemaking for
decommissioning reactors in the areas of emergency preparedness, indemnification, and
security. This study is one part of that basis.

The staff's assessment found that the frequency of spent fuel uncovery leading to a zirconium
fire at decommissioning SFPs is less than 5x10° per year (using the Lawrence Livermore
National Laboratory (LLNL) seismic hazard estimates (Ref. 1) for nuclear power plant sites)
when a utility follows certain industry commitments and certain of our recommendations. The
estimate drops to less than 1x10° per year if the EPRI site-specific seismic hazard estimates are
used. These frequencies are made up of contributors from a detailed risk assessment of
initiators (3.4x107 per year), both internal and external, and a quasi-probabilistic contribution
from seismic events (<5x10* per year using the LLNL hazard estimates or <6x107 per year
using the EPRI hazard estimates [Ref.2]) that have ground motions many times larger than
individual site design-basis earthquake ground motions (and higher uncertainty). It was also
determined that if these commitments and recommendations are ignored, the estimated
frequency of a zirconium fire could be significantly higher. Section 4 of this study discusses the
steps necessary to assure that a decommissioning plant operates within the bounds assumed in
the risk assessment. :

Previous NRC-sponsored studies have evaluated some severe accident scenarios for SFPs
at operating reactors that involved draining the SFP of its coolant and shielding water.
Because of the significant configuration and staffing differences between operating and
decommissioning plants, the staff performed this assessment to examine the risk associated
with decommissioning reactor SFPs.

First, the staff examined whether or not it was possible from a deterministic view point for a
zirconium cladding fire to occur. Zirconium fires were chosen as the key factor because
radionuclides require an energetic source to transport them offsite if they are to have a
significant health effect on local (first few miles outside the exclusion area) and more distant
populations. Deterministic evaluations in the staffs preliminary draft risk assessment indicated
that zirconium cladding fires could not be ruled out for loss of SFP cooling for fuel that has been
shut down and removed from an operating reactor within approximately 5 years. The

Appendix 2 A2-1


Attachment 3
Response to Questions Concerning Spent Fuel Pool

Seismic-Induced Failure Modes and Locations and the
Expected Level of Collateral Damage

by
Robert P. Kennedy
September 2000

1._ Introduction
This brief report responds to the following two questions from the NRC Staff:
What are the most likely spent fuel pool failure modes and locations?

L What is the expected level of collateral damage given a seismic
event necessary to fail the spent fuel pool?

The following responses are based upon my judgement without performing any
calculations.

2. Most Likely Spent Fuel Pool Failure Modes and Locations

Ref. 1 presents seismic fragility estimates for the Vermont Yankee (BWR) and
Robinson (PWR) spent fuel pools. These two fragility estimates are the only spent fuel
pool fragility estimates that I have seen. Therefore, my judgement is heavily based on the
results presented in Ref. 1.

For Vermont Yankee (BWR), Ref. 1 states that the critical failure mode for the
gross structural failure of the pool is an out-of-plane shear failure of the pool floor slab.
With this failure mode, the liner will be breached and a large crack will develop through
the concrete floor slab within a distance equal to the floor slab thickness from the pool
walls. Possibly the entire floor will drop out, but I think that such a gross failure is
unlikely. However, the concrete crack will be sufficiently large that the water in the pool
will quickly drain out.

Although not reported as the critical failure mode in Ref. 1, my judgement is that
for BWR pools, it is at least equally likely that the critical failure mode will be an out-of-
plane shear failure of one or more of the pool walls. With this failure mode, the liner will
be breached and a major concrete crack will form along the length of the wall within a
wall thickness distance from the top of the floor slab. Water will quickly drain out of the
pool. However, as much as 4-feet of water depth will likely remain within the pool.

A2B-23


licensing basis that.result in small increases in risk, such as relaxations in Emergency
Preparedness requirements.

Our conclusion in the draft final study was that, even though there are some differences in
source term and timing, scenarios involving an SFP zirconium fire may result in population
doses that are generally comparable to those expected from accident scenarios involving a
large early release at operating reactors, and therefore a PPG of 1x10° per year was
appropriate. The staff has reassessed these conclusions following the performance of
additional consequence calculations in Appendix 4A and 4B that took into account the
possibility of significant ruthenium release fractions. This assessment was undertaken to
address concems raised during review of the draft final study that large ruthenium releases
from a spent fuel fire could substantially increase both early and latent health effects, as well as
shift the controlling decision criteria from early fatalities to latent health effects because of the
combined effect of longer times for evacuation and longer ruthenium half life.

In reassessing the appropriateness of the 1x10° per year PPG as discussed below, the staff
contrasted the SFP risk for a licensee maintaining its facility at the PPG with the Commission's
Safety Goal Policy Statement. The Policy Statement expressed the Commission's policy
regarding the acceptable level of radiological risk from nuclear power plant operation as
follows:

. Individual members of the public should be provided a level of protection from the
consequences of nuclear power plant operation such that individuals bear no significant
additional risk to life and health

. Societal risks to life and health from nuclear power plant operation should be comparable
to or less than the risks of generating electricity by viable competing technologies and
should not be a significant addition to other societal risks.

The following quantitative health objectives (QHOs) are used in determining achievement of the
safety goals:

e The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities
that might result from reactor accidents should not exceed one-tenth of 1 percent
(0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which
members of the U.S. population are generally exposed.

. The risk to the population in the area near a nuclear power plant of cancer fatalities that
might result from nuclear power plant operation should not exceed one-tenth of 1 percent
(0.1 percent) of the sum of cancer fatality risks resulting from all other causes.

These QHOs have been translated into two numerical objectives as follows:

. The individual risk of a prompt fatality from all "other accidents to which members of the
U.S. population are generally exposed," such as fatal automobile accidents, is about
5x10“ per year. One-tenth of one percent of this figure implies that the individual risk of
prompt fatality from a reactor accident should be less than 5x10” per reactor year.

* “The sum of cancer fatality risks resulting from all other causes" for an individual is taken
to be the cancer fatality rate in the U.S. which is about 1 in 500 or 2x10° per year. One-
tenth of 1 percent of this implies that the risk of cancer to the population in the area near a
nuclear power plant because of its operation should be limited to 2x10* per reactor year.

Although the Policy Statement and related numerical objectives were developed to address the
tisk associated with power operation, is it reasonable to require that these objectives continue

Appendix 4C A4C-2


Bq. The reductions in prompt fatalities for uniform and Surry population densities are factors of 52 and
72, respectively. These reductions are more than proportional to the factor-of-four reduction in Cs-137
inventory, because of the combined effects of individual risk of early fatality and non-uniform population
density as discussed in the above analysis of the effect of ruthenium on offsite consequences.

For the cases with a ruthenium release fraction of one, the reduction in prompt fatalities is caused by the
reduction in the Ru-106 inventory which decreases from 5.77x10" Bq to 4.59x10" Bq in going from 3.5
cores to 1 core. This was confirmed by rerunning Case 34 with a Ru-106 inventory of 5.77x10" Bq. The
reductions in prompt fatalities for uniform and Surry population densities are factors of 1.30 and 1.89,
respectively. These reductions are nearly proportional to the factor of 1.26 reduction in the Ru-106
inventory. Again, deviations from being proportional are due to the combined effects of individual risk of
early fatality and non-uniform population density. Overall, the effect of reducing the number of
assemblies on prompt fatalities is less pronounced for the cases with a ruthenium release fraction of one,
in part, because the additional 2.5 cores has a small amount of Ru-106 (one year half-life) in comparison
with Cs-137 (30 year half-life). Finally, in all of the cases, the effect of reducing the amount of fuel
releasing fission products from 3.5 cores to one core is a modest decrease (20 to 40%) in societal dose
and cancer fatalities.

Plume-Related Parameters

The evaluation documented in Appendix 4 used the plume heat content associated with a large early
release for a reactor accident. The plume heat content for a spent fuel pool accident may be higher,
because (1) a spent fuel pool does not have a containment as a heat sink and (2) the heat of reaction for
zirconium oxidation is 85% higher in air than in steam. Also, the evaluation documented in Appendix 4
used the default values for the plume-spreading model in MACCS version 2.2, NUREG/CR-6244,
Probabilistic Accident Consequence Uncertainty Analysis, January 1995, provides improved values for
these parameters. .

Plume Heat Content:

The staff estimated that the complete oxidation in air (in a half hour) of the amount of zircalloy cladding in
a large BWR core would generate 256 MW. Subsequently, Sandia National Laboratories (SNL)
performed a more detailed assessment of the plume heat content for a spent fuel pool accident.* SNL
calculated that oxidation of 36% of the zircalloy cladding and fuel channels by the oxygen in the air flow
would heat up the accompanying nitrogen and the spent fuel to 2500 K. Once the spent fuel reaches
2500 K, it will degrade into a geometry in which continued exposure to air and, therefore, oxidation, will
be precluded. For a spent fuel pool accident involving the amount of fuel in a large BWR core, SNL
estimated the heat content of the nitrogen plume to be 43 MW. The SNL estimate was made by
subtracting (a) the energy absorbed by the spent fuel in heating up to 2500 K from (b) the energy
released by the oxidation of 36% of the zircalloy cladding and fuel channels.

The staff performed calculations with different plume heat contents to assess the sensitivity of the
consequences. The results of these calculations are shown in Table 12. Case 45, which used a
ruthenium release fraction of one, is shown in the second row of Table 12 and was the starting point for
these calculations. Case 45 used a plume heat content of 3.7 MW, which is associated with a large early
release for a reactor accident. Then, Cases 47 and 49 were run with plume heat contents of 83.0 MW.
and 256 MW, respectively. Increasing the plume heat content from 3.7 MW to 83.0 MW resulted in a
factor-of-two decrease in the early fatalities and no change in the long-term consequences. Increasing
the plume heat content from 83.0 MW to 256 MW resulted in a factor-of-three decrease in the early
fatalities and a small decrease in the long-term consequences.

-9-


APPENDIX 2B
STRUCTURAL INTEGRITY OF SPENT FUEL POOLS SUBJECT TO SEISMIC LOADS

1. INTRODUCTION

The staff's concern regarding seismic issues at spent fuel pools (SFPs) involves very large
earthquake ground motions that could catastrophically fail the SFP. Under this scenario, the
pool would suffer a significant breach, it would drain rapidly, and it will be incapable of being
refilled. This would lead to gradual cladding heat up, possibly followed by a zirconium cladding
fire. Attachment 1 to this appendix provides the checklist proposed by NEI and enhanced by the
staff to identify potential weaknesses and to assure adequate seismic capacity (at least 1.2g
peak spectral acceleration) at SFPs for decommissioning sites that wish to be granted
exemptions to EP, safeguards, and indemnification. Attachment 2 to this appendix provides the
analysis of the seismic failure probability of SFPs by the NRC’s consultant, Robert Kennedy, for
nuclear power plant sites based on a generic 1.2 g peak spectral acceleration high confidence,
with low probability of failure (HCLPF) value for SFPs. Attachment 3 to this appendix provides a
discussion of expected failure modes of SFPs because of very large ground motions and
describes the expected levels of collateral damage to the area surrounding the
decommissioning reactor site. The staff evaluated the frequency of large earthquake ground
motions at operating reactor sites, and these results are presented in Table 3 of Attachment 2 to
this appendix.

SFP structures at nuclear power plants are considered to be seismically robust. They are
constructed with thick reinforced concrete walls and slabs lined with stainless steel liners 1/8 to
1/4 inch thick. Pool walls are about 5 feet thick, and the pool floor slabs are around 4 feet thick.
The overall poo! dimensions are typically about 50 feet long by 40 feet wide and 55 to 60 feet
high. In boiling-water reactor (BWR) plants, the pool structures are located in the reactor
building at an elevation several stories above the ground. In pressurized-water reactor (PWR)
plants, the SFP structures are located outside the containment structure supported on the
ground or partially embedded in the ground. The location and supporting arrangement of the
pool structures influence their capacity to withstand seismic ground motion beyond their design
basis. The dimensions of the pool structure are generally derived from radiation shielding
considerations rather than seismic demand needs. Spent fuel structures at operating nuclear
power plants are able to withstand loads substantially beyond those for which they were
designed.

The Commission asked the staff to determine if there were a risk-informed basis for providing

exemptions from EP, safeguards, or indemnification for decommissioning plants and to provide
a technical basis for potential rule making. After this, the staff began to investigate the capacity
of SFPs to withstand large earthquake ground motions beyond the site’s seismic design bases.

‘Except for Dresden Unit 1 and Indian Point Unit 1, whose SFPs do not have any liner
plates. They were permanently shutdown more than 20 years ago, and no safety significant
degradation of the concrete pool structure has been reported.

Appendix 2B A2B-1


FAILURE OF MAKEUP
THRU AIRE PUMPS

Loom suc
i t 1
QOeRAT ER FAL TO. FAILS TORECOGNZE FALSTOSTART PW
INITIATE MAEUP BY USE NEED TOSTART PW Bie,
(BECTRICOR OES) Pues (QEDTRICGR DES)
HERRECOAWSTART-L HEPAWSTART
—————
FAILS TOREAPIRDESE. DES AND ELECTRIC
BLECTRICARE FIRE PROTECTIONPUNPS
PROTECTION PUMP FAL TOSTART ANDRUN
fo
1
FAILURE OF MAUP
THRUPRE PUMPS,
10.
1
f " 1
FALSTO REQOGNZE FAILS TOSTARTFW FIRE PROTECTION
NEED TOSTART FW BUMP PUMPS FAL TO.
PUMPS (BLECTRICOR DESSEL) PROVICE MAKEUP
oon 101 FART-LOt LB
DIESE. AND ELECTRIC FALSTO REPAR DIESE.
FREPROTECTION CTR
FAILTO START AND RUN. PROTECTION PUNP-
(2OHRS)
ve


readout in the control room (NEI commitment no. 5). There could also be indications or
alarms associated with pump flow and pressure, but no credit is taken here.

° The instrumentation is tested on a routine basis and maintained operable.

° Procedures are available to guide the operators in their response to off-normal conditions,
and the operators are trained on the use of these procedures (NEI commitment no. 2).

4.1.2.3 Quantification

Human Error Probabilities

The basic event HEP-DIAG-ALARM models operator failure to respond to an indication in the
control room and diagnose a loss of cooling event. Such an alarm would likely be the first
indication of trouble, so the operator would not be under any heightened state of alertness. On
the other hand, it is not likely that any other signals or alarms for any other conditions would be
present to distract the operator. The error rate is taken from THERP (Table 20-23).

Hardware Failure Probabilities

The value used for local faults leading to alarm channel failure (event SPC-LVL-LOP, 2.0E-3)
was estimated based on information in Reference 14. This event includes failure of
instrumentation and local electrical faults.

4.1.2.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-DIAG-ALARM 3.0E-4
SPC-LVL-LOP. 2.0E-3

4.1.3. Top Event IND — Other Indications of Loss of Cooling
4.1.3.1 Event Description and Timing

This top event models subsequent operator failures to recognize the loss of cooling during walk-
downs over multiple shifts. Indications available to the operators include: temperature readouts
in the control room (NEI commitment no. 5), local temperature measurements, and eventually,
increasing area temperature and humidity, low water level from boil-off, and local alarms.
Success for this event is defined as the operator recognizing the abnormal condition and
understanding the need to investigate its cause, leaving sufficient time to attempt to correct the
problem before the pool level drops below the SFP cooling system suction. The event is
modeled by fault tree LOC-IND.

4.1.3.2 Relevant Assumptions

. The loss of cooling may not be noticeable during the first two shifts but conditions are
assumed to be sufficient to trigger high temperature alarms locally and in the control room.

Appendix 2A A2A-14


A discussion of each of these safety principles and how they would continue to be satisfied at a
decommissioning plant is provided in the sections that follow. Since the application of this study
specifically relates to exemptions to a rule or a rule change for decommissioning plants, a
discussion of the first principle regarding current regulations is not necessary nor is it provided.

4.1.1 Increases in Risk

RG 1.174 states that when proposed changes result in an increase in core damage frequency
and/or risk, the increases should be small and consistent with the intent of the Commission's
safety goal policy statement.

The staff has evaluated the risks associated with SFP accidents and the impacts of potential
changes to regulatory requirements for decommissioning plants relative to applicable regulatory
guidance. Guidance on acceptable levels of (total) risk to the public from nuclear power plant
operation is provided in the Commission's safety goal policy statement (Ref. 3). Additional
guidance on the acceptable levels of risk increase from a change to the plant licensing basis is
provided in RG 1.174. The guidance contained in these documents is summarized below and
used in this study to evaluate the risks associated with SFP accidents and the impacts of
potential changes to regulatory requirements for decommissioning plants.

SFP Risk Relative to the Safety Goal Policy Statement

The “Policy Statement on Safety Goals for the Operation of Nuclear Power Plants,” issued in
1986, establishes goals that broadly define an acceptable level of radiological risk to the public
as a result of nuclear power plant operation. These goals are used generically to assess the
adequacy of current requirements and potential changes to the requirements. The Commission
established two qualitative safety goals that are supported by two quantitative objectives for use
in the regulatory decision-making process. The qualitative safety goals stipulate the following:

+ Individual members of the public should be provided a level of protection from the
consequences of nuclear power plant operation such that individuals bear no significant
additional risk to life and health.

* Societal risks to life and health from nuclear power plant operation should be comparable to
or less than the risks of generating electricity by viable competing technologies and should
not be a significant addition to other societal risks.

The following quantitative health objectives (QHOs) are used in determining achievement of the
safety goals:

+ The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities
that might result from reactor accidents should not exceed one-tenth of 1 percent
(0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which
members of the U.S. population are generally exposed.

4-2


Figure A2d-1 Rectangular Facility Effective Target Area Elements

Appendix 2D A2D-7


Only during the first several years and in the most severe events, such as severe seismic
events, heavy load drops, and other dynamic events, that cause the pool to fail, would the
accident progress so rapidly that emergency response measures might not be implemented in a
timely manner. The staff's risk study indicates that the frequency of such events is dominated
by earthquakes with a magnitude several times the safe shutdown earthquake (SSE). As
discussed in Section 3.7.2, for ground motion levels that correspond to SFP failure, emergency
planning would have marginal benefit because of extensive collateral damage to infrastructure
(e.g., power, communications, buildings, roads, and bridges). Emergency response action
would likely require substantial ad hoc action regardless of pre-planned actions in these events.

The next largest contribution is from cask drop sequences. The frequency of such events is low
in the staff's risk study (2x107 per year) due to implementation of IDCs and SDAs concerning
movement of heavy loads. Relaxations in EP requirements could result in some increase in the
tisk associated with these events for a limited time following shutdown (1 to 5 years in the staffs
analysis). However, the increase is a small fraction of the total risk from SFPs, as shown in
Section 3.7. For the remaining SFP accidents that were analyzed and lead to SFP fires (e.g.,
boildown sequences due to organizational failures), current emergency planning was assumed
to be ineffective or the frequencies of accidents, (e.g., aircraft impact) would be at least an order
of magnitude lower than for the cask drop accident. Thus, mitigation of these events would not
be risk significant.

With regard to defense-in-depth, the staff concludes:

+ Defense-in-depth for accident prevention and mitigation is provided by the robust design of
the SFP, the simple nature of pool support systems, and the long time available for taking
corrective action in response to system failures.

+ The substantial amount of time available for ad hoc offsite emergency response should
provide some level of defense-in-depth for consequence mitigation in SFP accidents.

+ Inthe large seismic events that dominate SFP risk, pre-planning for radiological accidents
would have marginal benefit due to extensive collateral damage offsite. Accordingly,
relaxations in EP requirements are not expected to substantially alter the outcome from
such a large seismic event.

+ There can be a tradeoff between the formality with which the elements of emergency
planning (procedures, training, performance of exercises) are treated and the increasing
safety margin as the fuel ages and the time available to respond gets longer.

4.1.3 Safety Margins

RG 1.174 states that the proposed change should maintain sufficient safety margins.

As discussed in Section 2, the safety margins associated with fuel in the SFP are much greater

than those associated with an operating reactor due to the low heat removal requirements and

long time frames available for recovery from off-normal events. Due to these larger margins, the
staff judges that the skid-mounted and other dedicated SFP cooling and inventory systems in

4-8


For Zero Dependence the probability of failure is Pog
Calculate P,,, using the appropriate values:

(1+( * )¥ = Task Failure Probability With Formal Dependence (P,,,)


October 26, 2000

MEMORANDUM TO: Gary M. Holahan, Director
Division of Systems Safety and Analysis
Office of Nuclear Reactor Regulation

FROM: Farouk Eltawila, Acting Director (Original signed by)
Division of Systems Analysis and Regulatory Effectiveness
Office of Nuclear Regulatory Research

SUBJECT: RADIOLOGICAL CONSEQUENCES OF SPENT FUEL POOL
ACCIDENTS OCCURRING UP TO 10 YEARS AFTER FINAL
REACTOR SHUTDOWN

As part of its effort to develop generic, risk-informed requirements for decommissioning, NRR
requested (Reference 1) that RES evaluate the offsite radiological consequences of beyond-
design-basis spent fuel pool accidents. In response to that user need, RES completed an in-
house analysis (References 2 and 3) using the MACCS code (Reference 4). The focus of that
work was estimation of consequences of accidents occurring between 30 days and 1 year after
final reactor shutdown. Recently, NRR requested (References 5 and 6) that RES extend the
consequence evaluation to accidents occurring up to 10 years after final shutdown.

RES performed the requested calculations using the release fractions in Table 1 and the fission
product inventories at 30 and 90 days and 1, 2, 5, and 10 years after final shutdown. The
release fractions in the first row of Table 1 are the sum of the in-vessel and ex-vessel release
fractions in NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,”
February 1995 (Reference 7). NUREG-1465 has received significant peer review and is
representative of a low pressure core-melt accident. The release fractions in the second row of
Table 1, other than those for ruthenium and fuel fines, also are from NUREG-1465. In this
case, the ruthenium release fraction is that for a volatile fission product, and the fuel fines
release fraction is that from the Chernobyl accident (Reference 8). Results of the RES
calculations for distances of 1, 10, and 50 miles are given in Tables 2 and 3.

Table 1 Fission Product Release Fractions

Source Release Fractions
Term
Xe,K |1 Cs Te Sr Ba Ru La Ce
r

NUREG- 1 7 75 31 12 12 .005 | .0052 | .0055
1465

NUREG- 1 75 15 31 12 12 75 035 035
1465

(modified)



Comment #13: How is common mode failure accounted for in the staff's risk analysis? How
confident are you of your ability to model and quantify common mode failures?

Response: The staff's risk analysis accounts for dependencies among the initiating events, the
equipment needed to mitigate the events, and also the operator actions needed for accident
mitigation. Initiating events that have the potential of simultaneously degrading mitigating
equipment or impeding operator actions are modeled in the construction of the event trees and
in the estimation of equipment failure rates and human failure probabilities. For example, for an
event where a fire is not extinguished within 20 minutes, it was assumed that the SFP cooling
system and the electric-driven firewater pumps are failed (either due to fire damage or due to
loss of the electrical supply to the plant). Therefore, no credit is taken for this equipment. In
addition, the estimation of the human error probability (for starting backup diese! pumps or for
off-site recovery) took into account a high level of operator stress, which increases the failure
probability.

Equipment hardware failure dependencies, usually referred to as common cause failures, have
also been modeled in the risk analysis. Since these failures have the potential for disabling
multiple trains of equipment at the same time, they can be large contributors to the risk. In the
staff's analysis, the only multiple train system modeled is the spent fuel pool cooling system. In
the fault tree model for this system, common cause failures are modeled for the cooling pumps,
the heat exchangers, and the discharge check valves. The modeling of dependent failures,
including common-cause hardware failures, in the staff's risk analysis is consistent with NRC
and industry guidelines. Based on the above, the staff has confidence that its modeling and
quantification of common mode failures is adequate.

Comment #14: NRC should set guidelines on how often fuel handlers make their rounds at
decommissioning facilities. This would help assure operator attentiveness.

Response: The staff agrees that, if fuel handlers make the rounds of the SFP and its
equipment on a frequent basis, the probability of the handlers detecting problems early is
greatly enhanced. To this end, SDA #2 states in part that walk-downs of the SFP systems will
be performed at least once per shift by the fuel handlers. The staff expects that these
assumptions will be translated into requirements or industry guidance during the rulemaking
process.

Comment #15: NRC should assure that the probability of failure of systems required to mitigate
the consequences of design bases and beyond design bases spent fuel pool events are
minimized.

Response: The need to have highly reliable systems to prevent or mitigate an accident is partly
a function of how rapidly the accident progresses and how serious its consequences are. If an
accident would result in serious consequences unless a rapid response were achieved, then
highly reliable systems and components are needed to prevent and/or mitigate the event. If the
accident is very slow in progressing or has benign consequences, the equipment designed to
prevent or mitigate it need not be as reliable. For SFPs at decommissioning plants, the large
volume of water above the spent fuel provides an inherent delay time before fuel can be
uncovered, except for two potential beyond design basis accidents which are discussed later.
This delay time (measured in days) allows for repair or replacement of equipment. If it were
impossible to repair or replace the equipment, inventory could be added to the pool to match
the boil-off rate. The industry has committed in IDC #4 to implement an off-site resource plan

Appendix 6 A6-5


available for fuel handler recovery from SFP events before initiation of a zirconium fire is
underestimated. However, since recoverable events such as small loss of inventory or loss of
power or pool cooling evolve very slowly, many days are generally available for recovery
whether the end point of the analysis is uncovery of the top of the fuel or complete fuel
uncovery. The extra time available (estimated to be in the tens of hours) as the water boils off
would not impact the very high probabilities of fuel handler recovery from these events, given
the industry decommissioning commitments (IDCs) and additional staff decommissioning
assumptions (SDAs) discussed in Sections 3.2 through 4.> A summary of the thermal-hydraulic
assessment is provided in Appendix 1A.

3.1 Basis and Findings of SFP Risk Assessment

To gather information on SFP design and operational characteristics for the preliminary risk
assessment for the June 1999 draft study, the staff visited four decommissioning plants to
ascertain what would be an appropriate model for decommissioning SFPs. The site visits
confirmed that the as-operated SFP cooling systems were different from those in operation
when the plants were in power operation. The operating plant pool cooling and makeup
systems generally have been removed and replaced with portable, skid-mounted pumps and
heat exchangers. In some cases there are redundant pumps. In most cases, physical
separation, barrier protection, and emergency onsite power sources are no longer maintained.
Modeling information for the PRA analysis was gathered from system walkdowns and
discussions with the decommissioning piant staff. Since limited information was collected for the
preliminary assessment on procedural and recovery activities and on the minimum configuration
for a decommissioning plant, a number of assumptions and bounding conditions were in the
June 1999 study. The preliminary results have been refined in this assessment, thanks to more
detailed information from industry on SFP design and operating characteristics for a
decommissioning plant and a number of IDCs that contribute to achieving low risk findings from
SFP incidents. The revised results also reflect improvements in the PRA model since
publication of the June 1999 and February 2000 studies.

The staff identified nine initiating event categories to investigate as part of the quantitative
assessment on SFP risk:

Loss of offsite power from plant centered and grid-related events
Loss of offsite power from events initiated by severe weather
Internal fire

Loss of pool cooling

Loss of coolant inventory

Seismic event

Cask drop

NQAARWON>

The staff notes that the assumption that no recovery occurs once the water level
reaches 3 feet above the fuel tends to obscure the distinction between two major types of
accidents: slow boildown or draindown events and rapid draindown events. In both types of
events, cooling would most likely be not by air but by water or steam. Also obscured is the
effect of partial draindown events on event timing (addressed in Section 2).

3-2


For the small leak, it is defined as the operator recognizing the abnormal condition and
understanding its cause in sufficient time to allow actions to prevent pool cooling from being lost.
Failure of this top event does not lead to fuel uncovery. This top event is represented by the
functional fault tree LOI-IND. Following an alarm, the operators would have in excess of 8 hours
before the water level would drop below the SFP cooling suction level. Therefore, for this event,
only one shift is credited for recognition.

For the large leak, success is defined as recognizing there is a leak in sufficient time to allow
make-up from alternate sources (fire water and offsite sources) before fuel uncovery. This top
event is represented by the basic event LOI-IND-L. Based on the success criterion, there are
many more opportunities for successive crews to recognize the need to take action. If the
leakage is in the SFP cooling system, the leak would be isolated automatically once the water
level drops below the SFP suction level. In this case, it would take more than 88 hrs (heatup
plus boil-off) for the water level to reach 3 ft above the top fuel and the event would be similar to
loss of SFP cooling. For the purpose of this analysis, it is assumed that leakage path is
assumed to be below SFP cooling system suction level. It is assumed that once the water level
drops 15 ft below normal pool level the leak is isolated automatically, and the inventory losses
would be only because of boil-off. Time needed to boil-off to 3 ft above the top fuel is estimated
to be 25 hours. Therefore, depending on the size of the leak and location and heatup rate, the
total time available for operator actions after the first alarm before the water level drops below
the SFP suction level to the 3 ft above the top of fuel would be more than 40 hrs. Furthermore,
the indications become increasingly more compelling; with a large leak it would be expected that
the water would be clearly visible, the level in the pool is obviously decreasing, and as the pool
boils the environment in the pool area becomes increasingly hot and humid. Because of these
very obvious physical changes, no dependence is assumed between the event IND and the
event CRA. This lack of dependence is however, contingent on the fact that the operating
crews perform walk-downs on a regular basis.

4.5.4.2 Relevant assumptions

¢ Operators have more than 40 hrs in the case of a large leak to take actions after the first
alarm before the water level drops to the 3 ft above the top of fuel.

. SFP water level indicator is provided in the control room e.g., camera or digital readout.
. SFP low-water level alarm (narrow range) is provided in the control room.
. System drawings are revised as needed to reflect current plant configuration.

¢  Procedure/guidance exist for the operators to recognize and respond to indications of loss
of inventory, and they are trained in the use of these procedures (NEI commitment no. 2).

« Water level measurement stick with clear marking is installed in the pool at a location that
is easy to observe

* Operators are required to make a round per shift and document walk-downs in a log

Appendix 2A A2A-56


Table A2c-2 Summary of NUREG-0612 Heavy Loads Evaluation (for cask drop) with New
1990s Navy Crane Data Values and WIPP Rigging HEP Method

Event Description Units {High Low Mean
INO [Base range of failure of handling system year 1.5e-04[ 1.0e-05| _5.4e-05
(Crane Failure
Ft Fraction of load hangup events (new 1990s Navy data) i 0.14 0.14 0.14
ICF14 |Operator error leading to load hangup (NO*F1)) year 2.0e-05| 1.4e-06} _7.4e-06
/ideman
\CF12 Failure of the overload device id 1.0e-02] 1.0e-03]  4.0e-03
CF1 Load hangup event (CF11*CF12) Vyear 2.0e-07| 1.4e-09| _3.0e-08
F2 Fraction of component failure events (new 1990s Navy data) i 0.61 0.61 0.61
cF21 Faiture_of single component with a backup (NO*F2) year 9.1e-05| 6.1e-06| _3.3e-05
/deman
|\CF22 Failure of backup component given CF21 id 1.0e-01} 1.0e-02] 4.0¢-02
ICF2 Failure because of random component failure (CF21*CF22) ar 9.1e-06] 6.1e-08]  1.3e-06
FS. Fraction of two-blocking events (new 1990s Navy data) i— 0.05, 0.05, 0.05
\CF31 |Operator error leading to Two-blocking (NO*F3) year 6.8e-06] 4.5e-07] 2.5e-06
fderman
|CF32 Failure of lower limit switch id 1.0e-02| 1.0e-03] _4.0e-03
/deman
|CF33 Failure of upper limit switch jd 1,0e-01] 1.0e-02| 4.0e-02
|CF3 [Two-blocking event (CF31*CF32*CF33) year 6.8e-09| 4.5e-12|  4.0e-10
F4 Fraction of single component failure (new 1990s Navy data) — 0.04 0.01 0.01
ideman
Far [Credit for NUREG-0554 id 0.10 0.10 0.10
CFS Failure of component that doesn't have backup (NO*F4*F4') fyear 2.2e-07| 1.5e-08{ 8.1e-08
CRANE __|Failure of crane (CF1+CF2+CF3+CF4) year 9.5e-06| 7.7e-08]  1.4e-06
D1 Lifts per year leading to drop (100 lifts per year, drops from non-rigging) |No. 3 3 3
CF Failure of crane leading to load drop (CRANE*D1) year 2.9e-05| 2.3e-07| 4.4e-06
Rigging failure - Based on WIPP method
FS [Fraction of improper rigging events (new 1990s Navy data) i— 0.21 0.21. 0.21
CR11 Failure because of improper rigging, mean from WIPP study ear 8.7e-07] 8.7e-07| 8.7e-07
CR12 Failure of redundant/atternate rigging INA
RIGGING _|Failure because of improper rigging (CR11) year 8.7e-07| 8.7e-07| 8.72-07
D2 [Lifts per year leading to drop (100 lifts per year, drops from rigging) INo. 6 6 6
CR Failure of rigging leading to a load drop (RIGGING*D2) liyear 5.3e-06| 5.3e-06| 5.3¢-06
Failure of heavy load (crane and rigging) system
FHLS (CRANE+RIGGING) year 1.0e-05| 9.5e-07|  2.3e-06
ICFCR _ | Total failures (crane and rigging) leading to aload drop (CF+CR) _|fyear 340-05] 5.5¢-06| 9.6¢.06
|Loss-of-inventory for a single-failure-proof crane
RF Fraction of year over which a release may occur = 1.00 1.00 1.00
iP Fraction of path near/over pool a 0.25 0.05 0.13
Pr Fraction of path critical for load drop i 0.25 0.10 0.16
LOI-S (CFCR) * P* P'* RE year 2.1e-06} 2.8¢-08] 2.0e-07
Loss-of-inventory for a non single-failure-proof crane
CFCRNO.
IN {Total failures leading to a dropped load (est. from NUREG-0612) No. 750-05] 1.0e-07| 2.1e-05
RE Fraction of year over which a release may occur > 1.00 1,00 1,00
[LOI-N (CFCRNON) * P * P'* RF year 7.5e-05| 1.0e-07| _ 2.1¢-05
Risk reduction for a single-failure-proof crane (LOI-N /LOI-S) i 35. 4 104

A2C-7



Federal Regulations, Appendix A, General Design Criteria, which require safe handling and
storage under normal and accident conditions.

Regarding degradation, type 304 stainless steel material, which is used for the spent fuel racks,
is susceptible to stress corrosion cracking in oxygenated water environment at relatively high
temperature conditions. At the temperature levels that exist in the spent fuel pools, stress
corrosion cracking of the spent fuel racks made of stainless steel is not a concern, and there
has been no report of any actual incidence of stress corrosion cracking of spent fuel racks.

The stagnant, borated condition of the spent fuel pool water is not a significant factor in
inducing stress corrosion cracking of the racks. Most spent fuel assemblies are clad with
zirconium and are not known to be susceptible to stress corrosion cracking.

Comment #61: A significant seismic event which damages and drains the SFP is also likely to
wreak havoc upon the local infrastructure. How has NRC considered the availability of local
resources as identified by IDC #2, #3, and #4 should the local infrastructure be destroyed?

Response: Seismic capacity of spent fuel structures against catastrophic failures, such that a
very rapid loss of water can be assumed, is substantially above the safe shutdown earthquake
levels of the spent fuel pools. Consequently, high ground motion levels are necessary to
initiate failures. The response by local, state, or national authorities needed at the spent fuel
pool site will depend on the actual or potential damage to the spent fuel pool. The most likely
damage to the spent fuel pool and support systems would be to the support systems that
provide cooling to the pool. The large inventory of water above the spent fuel should provide
adequate time (it would take about a week without pool cooling before boiling would occur) for
repairing or bringing in replacement pumps and heat exchangers. If the local infrastructure was
damaged by a seismic event such that the prearranged off-site response could not occur, the
industry commitments provide a good foundation for an ad hoc response.

Comment #62: For all central and eastern U.S. nuclear power plant sites and for some western
U.S. nuclear power plant sites, all that is necessary to have an adequately safe spent fuel pool
with respect to seismic-induced risk is for the pool to meet the requirements of the seismic
checklist. Several western U.S. sites may need to demonstrate a high confidence with low
probability of failure (HCLPF) of 2 X SSE.

Response: The staff agrees that, for most sites throughout the U.S., meeting the enhanced
seismic checklist (Appendix 2D) is sufficient to demonstrate acceptable seismic risk for
decommissioning spent fuel pools. However, if a site does not pass the seismic checklist, a
plant-specific seismic risk evaluation of the spent fuel pool could be performed. The staff is no
longer recommending using 2 or 3 times the SSE as a criterion. The staff's re-evaluation is
included in this report.

Comment #63: The value of three times the SSE for the SFP HCLPF should not be a hard and
fast acceptance criteria, since this is only a screening criteria.
Response: The staff is no longer recommending the use of 2 or 3 times the SSE as a criterion.

This report provides the staff's re-evaluation of the SFP accident risk at decommissioning
plants and the criteria necessary to demonstrate low risk.

Appendix 6 AG-21


reduce the additional time gained to reach 1200 °C to a relatively small amount. Selection of a
temperature criterion for long-term fuel pool integrity needs to consider that ruthenium release
rates, in air, become significant at approximately 600-800 °C, based on the data of Parker et al.
(Ref. 7).

Selection of an acceptance criterion for precluding significant offsite release after roughly

5 years, should also consider that ruthenium with a 1 year half life will be substantially decayed
and that at 5 years cesium (and perhaps fuel fines such as plutonium) will dominate the dose
calculation. For these reasons RES believes that the long-term viability of the pool in a
completely drained condition (air environment), if it concerns time periods of approximately

5 years, pool degradation should be assessed for a temperature of approximately 800 °C.
Again, an analysis needs to be performed to demonstrate that at that temperature an equilibrium
condition can be established. While this would result in an offsite release, there would be
substantial time available to take corrective action after a 5 year decay time for the most recently
loaded fuel. If shorter decay time periods are proposed for achieving the long-term equilibrium
temperature criterion, then the impact of ruthenium releases would dictate reconsideration of
this value.

4. SUMMARY

In summary, we conclude that for assessing the onset of fission product release under transient
conditions (to establish the critical decay time for determining availability of 10 hours to
evacuate) it is acceptable to use a temperature of 900 °C if fuel and cladding oxidation occurs in
air. If steam kinetics dominate the transient heatup case, as it would in many boildown and
draindown scenarios, then a suitable temperature criterion would be around 1200 °C. For
establishing long-term equilibrium conditions for fuel pool integrity during SFP accidents which
preclude significant fission product release it is necessary to limit temperatures to values of

600 °C to 800 °C. If the critical decay time is sufficiently long (>5 yrs) that ruthenium inventories
have substantially decayed then it would be appropriate to consider the use of a higher
temperature, 800 °C, otherwise fission product releases should be assumed to commence at
600 °C. These cases are marked by substantial time for corrective action to restore cooling and
prevent smaller gap type releases associated with early cladding failures. A tabulated summary
of the suggested criteria is listed below.

Precluding Large Precluding Large
Adequacy of 10 hrs Release Release
for Evacuation Fuel <5yrs Fuel >5yrs
Dominant Air
Environment 900 °C 600 °C. 800 °C
Dominant Steam
Environment 1200 °C N/A NIA

The degradation of fuel during SFP accidents is an area of uncertainty since most research on
severe fuel degradation has focused on reactor accidents in steam environments. Because of
this uncertainty, we have tended to rely on the selection of conservative criterion for predicting

Appendix 1B A1B-6


3.3.1 Internal and External Initiator Frequency of Spent Fuel Pool Uncovery

The results for the initiators that were assessed quantitatively are shown in Table 3.1. This table
gives the fuel uncovery frequency for each accident initiator. The frequencies are point
estimates because point estimates were used for the input parameters. For the most part, these
input parameter values are the mean values of the probability distributions that would be used in
a calculation to propagate parameter uncertainty. Because the systems are very simple and
needs little support, the point estimates therefore reasonably correlate to the mean values that
would be obtained from a full propagation of parameter uncertainty. Due to the large margin
between the loss of cooling and inventory sequence frequencies and the pool performance
guideline, this propagation was judged to be unnecessary (see Section 5.1 of Appendix 2A for
further discussion of uncertainties).

Both the EPRI and LLNL hazard estimates at reactor sites were developed as best estimates
and are considered valid by the NRC. Furthermore, because both sets of curves are based
upon expert opinion and extrapolation in the range of interest, there is no technical basis for
excluding consideration of either set of estimates. The mean frequency shown does not
consider Western U.S. sites (e.g., Diablo Canyon, San Onofre, and WNP-2).

The results in Table 3.1 show the estimated frequency of a zirconium fire of fuel that has a
decay time of 1 year. In characterizing the risk of seismically-induced SFP accidents for the
population of sites, the staff has displayed results based on both the LLNL and EPRI hazard
estimates, and has used an accident frequency corresponding to the mean value for the
respective distributions, i.e., a frequency of 2x10°* per year to reflect the use of LLNL hazard
estimates and a frequency of 2x10” per year to reflect use of the EPRI hazard estimates. Use
of the mean value facilitates comparisons with the Commission’s Safety Goals and QHOs.

Fire frequencies for all initiators range from about 6x10” per year to about 2X10* per year
(depending on the seismic hazard estimates used), with the dominant contribution being from a
severe seismic event. Plant-specific frequency estimates in some cases could be as much as
an order of magnitude higher or lower because of the seismic hazard at the plant site. The
mean value bounds about 70 percent of the sites for either the LLNL or the EPRI cases. A more
detailed characterization of the seismic risk is given in Section 3.5.1 and Appendix 2B. The
frequency of a zirconium fire is dominated by seismic events when the seismic hazard frequency
is based on the LLNL estimate. Cask drop and boildown sequences become important
contributors when seismic hazard frequency is based on the EPRI estimate. As a result, even
though the seismic event frequency based on the EPRI estimate is an order of magnitude lower
than the LLNL estimate, only a factor of four reduction in total frequency is realized with the use
of the EPRI estimate since the nonseismic sequences become more important. In Section 3.4.7
the staff discusses the expected fuel uncovery frequencies for fuel that has decayed a few
months, 2 years, 5 years, and 10 years.

In conjunction with the frequency of the uncovery of the spent fuel, it is important to know the
time it takes the fuel to heat up once it has been uncovered fully or partially. Figures 2.1 and 2.2
‘in Section 2 show the time needed with and without air circulation to heat up the fuel from 30 °C
to 900 °C (the temperature at which zirconium oxidation is postulated to become runaway
oxidation and at which fission products are expected to be expelled from the fuel and cladding).

37


is difficult for offsite resources to reach the facility or if regional resources are engaged in other
efforts, then it would also be unlikely that the population in the area would be effectively notified
and/or evacuated under these conditions. For sequences other than loss of off-site power due
to severe weather, the dominant reason that recovery is not provided in the failure paths is a
general breakdown in the overall facility organization. The failure to acquire offsite resources
also implies a failure to contact regional authorities and declare an emergency when the SFP
level drops below the proceduralized limit in these sequences. Accordingly, the consequences
for boildown sequences are based on results for the late evacuation case (Tables 3.7-1 and
3.7-2). This same reasoning is applied for cases with and without EP relaxations and for all
times after shutdown. The net effect is that EP, as well as relaxations in EP, do not impact the
risk associated with those boildown sequences that proceed to spent fuel uncovery.

Rapid Draindown Due to Seismic Events

Given the robust structural design of SFPs, it is expected that a seismic event with peak spectral
acceleration several times larger than the safe shutdown earthquake (SSE) would be required to
produce catastrophic failure of the structure. The estimated frequency of events of this
magnitude differs greatly among experts and is driven by modeling uncertainties. The estimated
frequency of seismic events sufficiently large to result in structural failure of the SFP is given in
Table 3.7-4 and is based on the LLNL and EPRI seismic hazard estimates.

Both the LLNL and EPRI hazard estimates were developed as best estimates and are
considered valid by the NRC. Furthermore, because both sets of curves are based upon data
extrapolation and expert opinion, there is no technical basis for excluding consideration of either
set.

Using the LLNL hazard estimates, a return frequency equivalent to the pool performance
guideline (1x10° per year) for a 1.2g peak spectral acceleration (PSA) ground motion bounds all
but four sites (one Central and Eastern and three Western U.S. sites). The frequency for the
remaining sites falls in the range of less than 7x10° per year to 9x10* per year. The majority
(45 sites) have hazard estimates (for a 1.2 PSA ground motion) near 1x10° per year and

20 sites fall below 6x10” per year. The mean value for the population of plants is approximately
2x10° per year.

lf EPRI hazard estimates were used, only one site would have an estimate that exceeds 1x10°
per year (excluding Western sites)."* Ten sites are near 5x10” per year, and the remaining

49 sites analyzed by EPRI have estimates less than 3x10” per year, with half of these sites

(25 sites) estimated at less than 7x10* per year. The mean value for the population of plants is
approximately 2x10” per year.

In characterizing the risk of seismically induced SFP accidents for the population of sites, the
staff has displayed results based on both the LLNL and the EPRI hazard estimates, and has
used an accident frequency corresponding to the mean value for the respective distributions,

*SEPRI seismic estimates were not developed for all sites east of the Rocky Mountains.
Six sites have LLNL but no EPRI hazard estimates.

3-36


outset RES acknowledges that an ignition temperature, or more precisely in this case a
temperature for incipient temperature escalation is dependent on heat generation and losses
which in turn is dependent on system geometry and configuration. In fact, much of the data on
oxidation is produced in isothermal tests up to near the melting temperature of zirconium. In
examining an appropriate criterion, it is useful to consider the range of available data including
core degradation testing in steam environments, since it is likely that many SFP accidents may
involve some initial period during which steam oxidation kinetics controls the initial oxidation,
heatup, and release of fission products. In various experimental programs around the world
(e.g., PBF-SFD, ACRR, CORA, NSRR, PHEBUS and QUENCH) repeatable phenomena have
been observed for the early phase of core degradation (in steam) which proceeds initially at
temperature increase rates associated with decay heat (at levels characteristic of reactor
accidents) until cladding oxidation becomes dominant and a more rapid temperature escalation
occurs. The point at which the escalation occurs, which does vary between tests, has been
attributed to heat losses (Ref. 2) characteristic of the facility and to phase changes of ZrO, over
a temperature range. The threshold at which temperature escalation occurred has been
reported to vary from approximately 1100 °C to 1600 °C. In a CORA test performed with a lower
initial heatup rate (to simulate reduced decay heat during shutdown conditions) it was reported
that uncontrolled temperature escalation did not occur, raising the prospect that heating rate
may be a factor. (This is probably because of the formation of a thicker oxidation layer built up
over the protracted time at lower temperature such that when higher temperatures are attained,
the thicker scale results in a lower oxidation rate relative to a thinner scale at the same
temperature.) In more recent QUENCH tests (Quench 04 and 05) the effect of preoxidation was
evaluated for its effect on hydrogen generation and temperature escalation. In Quench 04
temperature escalation was reported to occur at 1300 °C; in Quench 05 with approximately
200um preoxidation temperature escalation was reported to be delayed until the fuel rod
temperature reached 1620 °C.

Because of interest in air ingression phenomena for reactor accidents, recent severe accident
research has also examined oxidation in air environments. Publication of results from the
DRESSMAN and CODEX test programs (Ref. 3) has provided much of the transient data on fuel
rod and rod bundle behavior for air kinetics as well as data on fuel oxidation and volatility. Early
studies of zirconium oxidation in air (Refs. 4 and 5) were performed by comparing isothermal
oxidation and scaling of fresh samples to determine the influence of different atmospheres and
materials as well as to examine potential for fire hazards. The general observation is that, at
least at higher temperatures(>1000 °C), the oxidation rate is higher in air than in steam.

Another observation of the early studies was, under the same conditions, oxidation in an air
environment produced an oxide layer or scale less protective than that for steam owing to the
possible instability of a nitride layer beneath the outer oxide layer leading to scale cracking and
a breakaway in the oxidation rate. The onset of this breakaway in the oxidation rate occurred at
about 800 °C after a time period of 10 hours in the studies performed by Evans et al (Ref. 4) and
after a period of approximately 4 hours in studies by Leistikow (Ref. 9). The Leistikow studies
were performed on fresh cladding, however, and it is expected that breakaway would occur after
a longer time delay with preoxidized cladding. As breakaway oxidation occurs the oxidation
behavior observed no longer reflects a parabolic rate dependence but takes on a linear rate
dependence. Also, at lower temperatures the kinetics of reaction indicate near cubic rate
dependence thus the representation of the oxidation behavior at both high and low temperatures
with a parabolic rate dependence may introduce unnecessary simplification and an

Appendix 1B A1B-2



The second evaluation was presented in a paper by K. Jamali [Ref. 4] in which additional facility
evaluations were summarized. For the Seabrook Nuclear Power Station, Jamali’s application of
the DOE effective aircraft target area model to the final safety analysis report (FSAR) data
resulted in an impact frequency 2.4x10° per year. The Millstone Unit 3 plant area was reported
as 9.5x10° square miles and the FSAR aircraft crash frequency as 1.6x10° per year. Jamali
applied the DOE effective aircraft target area mode! to information in the Millstone Unit 3 FSAR.
Jamali reported an impact frequency of 2.7x10* per year, using the areas published in the
FSAR and 2.3x10° per year, and using the effective area calculated the effective aircraft target
area model.

When the generic DOE data in Table A2d-1 were used (for a 514 x 514 x 100-foot site), the
estimated impact frequency range was 6.3x10° to 2.9x10° per year, with an average of 1.9x10*
per year, for the point target area model. The effective aircraft target area model gave an
estimated range of 3.1x10* to 2.4x10* per year, with an average of 1.6x10° per year.

A site-specific evaluation for Three Mile Island Units 1 and 2 was documented in
NUREG/CR-5042 [Ref. 5]. The NUREG estimated the aircraft crash frequency to be 2.3x10+
accidents per year, about the same value as would be predicted with the DOE data set for the
maximum crash rate for a site area of 0.01 square miles.

NUREG/CR-5042 summarized a study of a power plant response to aviation accidents. The
results are given in Table A2d-4. The probability of the penetration of an aircraft through
reinforced concrete was taken from that study.

Based on comparing these plant-specific aircraft crash evaluations with the staff's generic
evaluation, there were no significant differences between the results from the DOE model
whether generic data were used to provide a range of aircraft crash hit frequencies or whether
plant-specific evaluations were performed.

3. ESTIMATED FREQUENCIES OF SIGNIFICANT SFP DAMAGE

The frequency for significant PWR SFP damage resulting from a direct hit was estimated based
on the point target model for a 100 x 50-foot pool with a conditional probability of 0.45 (large
aircraft penetrating 5-ft of reinforced concrete) that the crash resulted in significant damage.
This value (i.e., 0.45) is an interpolation from a table in NUREG/CR-0542 reproduced in Table
A2d-4. If 1-of-2 aircraft are large and 1-of-2 crashes result in spent fuel uncovery, then the
estimated range is 1.3x10"' to 6.0x10° per year. The average frequency was estimated to be
4.1x10° per year.

The mean frequency for significant BWR SFP damage resulting from a direct hit was estimated
to be the same as that for the PWR, 2.9x10° per year.

4. SUPPORT SYSTEM UNAVAILABILITY

The frequency for loss of a support system (e.g., power supply, heat exchanger, or makeup
water supply) was estimated based on the DOE model, including wing and skid area, for a 400 x
200 x 30-foot area with a conditional probability of 0.01 that one of these systems is hit. The

Appendix 2D A2D-3


embedded in the ground. The location and supporting arrangement of the pool structures affect
their capacity to withstand seismic ground motion beyond their design basis. The dimensions of
the pool structure are generally derived from radiation shielding considerations rather than
seismic demand needs. Spent fuel structures at nuclear power plants are able to withstand
loads substantially beyond those for which they were designed.

To evaluate the risk from a seismic event at an SFP, one needs to know both the likelihood of
seismic ground motion at various acceleration levels (i.e., seismic hazard) and the conditional
probability that a structure, system, or component (SSC) will fail at a given acceleration level
(i.e., the fragility of the SSC). These can be convolved mathematically to arrive at the likelinood
that the SFP will fail from a seismic event. In evaluating the effect of seismic events on SFPs, it
became apparent that although information was available on seismic hazard for nuclear power
plant sites, the staff did not have fragility analyses of the pools, nor generally did licensees. The
staff recognized that many of the SFPs and the buildings housing them were designed by
different architect-engineers. Additionally, the pools were built to different standards as the rules
changed over the years.

To compensate for the lack of knowledge of the capacity of the SFPs, the staff proposed the use
of a seismic checklist during stakeholder interactions, and in a letter dated August 18, 1999, NEI
proposed a checklist that could be used to show an SFP would retain its structural integrity at a
peak spectral acceleration of about 1.2 g. This value was chosen, in part, because existing
databases that could be used in conjunction with the checklist only go up to 1.2 g peak spectral
acceleration. The checklist was reviewed and enhanced by the staff (see Appendix 2B). The
checklist includes elements to assure there are no weaknesses in the design or construction or
any service-induced degradation of the pools that would make them vuinerable to failure during
earthquake ground motions that exceed their design-basis ground motion but are less than the
1.2 g peak spectral acceleration. The staff used a simplified, but slightly conservative method to
estimate the annual probability of a zirconium fire due to seismic events and site-specific seismic
hazard estimates (see Appendix 2B, Attachment 2). These calculations resulted in a range of
frequencies from less than 1x10° per year to over 1x10° per year, depending on the site and the
seismic estimates used.

Figures 3.2 and 3.3 show the estimated annual probabilities of a zirconium fire from a seismic
event in ascending order. Figure 3.2 shows the results of convolving the site-specific LLNL
seismic hazard estimates (from NUREG-1488, “Revised Livermore Seismic Hazard Estimates
for 69 Nuclear Power Plant Sites East of the Rocky Mountains,” P. Sobel, October 1993) with
the generic SFP fragility analysis, and Figure 3.3 shows the results of convolving the EPRI site-
specific seismic hazard estimates (Ref. 10) in a similar manner.'® Note that the order of the sites
differs somewhat In the EPRI and LLNL estimates. These figures show that for the zirconium
fire frequencies using the LLNL estimates, the annual probabilities for most site clusters just
above 1x10 per year. The mean failure probability for the sites analyzed by LLNL is about
2x10* per year. This value bounds 70 percent of the sites using the LLNL curves. For the EPRI

*° At higher accelerations, especially for plant sites east of the Rocky Mountains, there is
great modeling uncertainty about the ground motions, return periods, and the possibility of
cutoff. There is virtually no data at these acceleration levels.

3-19


walk-downs are performed on a regular, (once per shift) basis is important to compensate for
potential failures to the instrumentation monitoring the status of the pool, the assumption that the
procedures and/or training are explicit in giving guidance on the capability of the fuel poo!
makeup system, and when it becomes essential to supplement with alternate higher volume
sources, the assumption that the procedures and training are sufficiently clear in giving
guidance on early preparation for using the alternate makeup sources, are crucial to
establishing the low frequency. NEI commitments 6, 7 and 9 have been credited with lowering
the initiating event frequency.

Appendix 2A A2A-64


Response: The specific reductions in the areas mentioned is a subject that is beyond the intent
of this study and will be determined during the rulemaking process. Generally speaking, it is
anticipated that on-site manpower could be reduced early in the decommissioning process
provided adequate personnel are available to provide emergency response duties. Off-site
manpower needs, equipment, communication, alarms, notifications, plans, and planning areas,
would be relaxed consistent with the relaxation of requirements for off-site emergency planning.
The consideration of the use of KI would not be necessary when iodine releases are no longer
a concern.

Comments #92 and 93: (A) It is difficult to figure out how this effort fits into the overall big
picture of what the NRC is doing on decommissioning. (B) A commenter asked the staff to
“look at all of the activities that happen during decommissioning when developing regulations,
not just a narrow view of the spent fuel pool.”

Response: The focus of the decommissioning spent fuel pool risk study was intentionally
limited to address potential severe accidents associated only with spent fuel. An additional
rulemaking effort, termed the regulatory improvement initiative, is planned by the NRC and will
include a comprehensive look at all decommissioning regulations to determine if any additional
changes are required. An overall assessment of decommissioning issues and other activities
that take place at decommissioning sites will be addressed during this subsequent effort.

Comments #94 and 95: (A) A commenter stated that he was confused on the way Part 50 is
being applied in places where Part 72 might be more applicable. (B) Why does the NRC apply
Part 50 (reactor) regulations to decommissioning reactors when the rules in Part 72 for storage
of high-level waste are more clearly outlined? Part 50 regulations are not appropriate for long-
term storage of high-level waste.

Response: Although 10 CFR Part 50 was developed with the operating power reactors in
mind, many of the requirements still apply to decommissioning power reactors. The NRC
believes that the 10 CFR Part 50 regulations applicable to decommissioning reactors are
sufficient to assure public health and safety. The Part 50 license allows for safe storage of
spent fuel in a spent fuel pool during operation and the staff believes that license remains
adequate for spent fuel pool storage during decommissioning. The staff does not require a Part
50 licensee to obtain a Part 72 license for spent fuel storage in a spent fuel pool. All reactor
decommissioning activities will remain under the Part 50 license until the decommissioning is
completed and the license is formally terminated. When a licensee chooses to store spent fuel
in an independent spent fuel storage installation, then the requirements of Part 72 license and
regulations will be applicable.

In SECY-99-168, dated June 30, 1999, the NRC staff proposed to the Commission that all NRC
regulations under Title 10 be reviewed and modified as necessary to ensure proper applicability
to decommissioning. At the direction of the Commission, the staff is currently performing a
comprehensive review of all applicable NRC regulations that may need modification to more
effectively address decommissioning reactors.

Appendix 6 A6-31


4.2.4.3 Quantification

Failure of fire suppression is represented by basic event HEP-RES-FIRE. The modeling of fire
growth and propagation and the determination of the effects of a fire on equipment in a room
would optimally take into account the combustible loading in the room, the presence of
intervening combustibles, the room size and geometry, and other characteristics such as
ventilation rates and the presence of openings in the room. Because detailed inputs such as
these are not applicable for a generic study such as this, fire growth and propagation was
determined based on best estimate assumptions. It is assumed that the operator has

20 minutes to suppress the fire. Otherwise, it is assumed that SFP cooling will be lost (due
either to damage of SFPC equipment, or to the plant’s power supply system).

HEP-RES-FIRE was modeled using THERP. Because of the level of uncertainty about the size
of the fire, its location, and when it is discovered, the approach taken was to model this error as
a dynamic task requiring a higher level of human interaction, including keeping track of multiple
functions. In addition little experience in fighting fires was assumed. Table 20-16 in THERP
provides modifications of estimated HEPs for the effects of stress and experience. Using the
performance shaping factors of extremely high stress (as fighting a fire would be), a dynamic
task, and an operator experienced in fighting fires, this table provides an HEP of 2.5E-1.

Notes: (1) It can be argued that damage time (to disable the SFP cooling function) could be
in excess of 20 minutes because typical SFP facilities are relatively large and
because equipment within such facilities is usually spread out. However, in this
analysis, the SFP pumps are assumed to be located in the same general vicinity
with no fire barriers between them.

(2) Scenarios can be postulated where the fire damage state is less severe than that
described above (e.g., fire damage to the running cooling pump, with the other
pump undamaged, and with offsite power available). These scenarios can be
subsumed into the "Loss of Cooling" event, and SFP cooling "recovery" in these
cases would be by use of the undamaged pump train.

4.2.4.4 Basic Event Probability

Basic Event Basic Event Probability
HEP-RES-FIRE 2.5E-1

4.2.5 Top Event OMK - Operator Recovery Using Onsite Sources
4.2.5.1 Event Description and Timing

At this point in the event tree, the SFP cooling has been lost as a result of the fire, and the
operators are unable to restore the cooling system. Also, the fire has damaged the electrical
system such that the motor-driven firewater pump is unavailable. If no actions are taken, SFP
water level would drop to 3 ft above the top of fuel in 131 hours from the time the loss of SFP
cooling occurred. This event represents failure of the operators to start the diesel-driven

Appendix 2A A2A-29


8. Aircraft impact
9. Tornado missile

In addition, a qualitative risk perspective was developed for inadvertent criticality in the SFP (see
Section 3.6). The risk model, as developed by the staff and revised after a quality review by
Idaho National Engineering & Environmental Laboratory (INEEL), is provided in Appendix 2A.
Appendix 2A also includes the modeling details for the heavy load drop, aircraft impact, seismic,
and tornado missile assessments. Input and comments from stakeholders were also utilized in
updating the June 1999 and February 2000 risk models.

3.2 Characteristics of SFP Design and Operations for a Decommissioning Plant

Based on information gathered from the site visits and interactions with NEI and other
stakeholders, the staff modeled the spent fuel pool cooling (SFPC) system (see Figure 3.1) as
being located in the SFP area and consisting of motor-driven pumps, a heat exchanger, an
ultimate heat sink, a makeup tank, a filtration system, and isolation valves. Coolant is drawn
from the SFP by one of the two pumps, passed through the heat exchanger, and returned to the
pool. One of the two pumps on the secondary side of the heat exchanger rejects the heat to the
ultimate heat sink. A small amount of water is diverted to the filtration process and is returned
into the discharge line. A manually operated makeup system (with a limited volumetric flow rate)
supplements the small losses due to evaporation. During a prolonged loss of the SFPC system
or a loss of inventory, inventory can be made up using the firewater system, if needed. Two
firewater pumps, one motor-driven (electric) and one diesel-driven, provide firewater in the SFP
area. There is a firewater hose station in the SFP area. The firewater pumps are in a separate
structure.


Table § Number of Prompt Fatalities in Each Radial Ring for Cases with Late Evacuation

Distance Number of early fatalities with Surry Number of early fatalities with uniform
(km) population density population density
Base Case, Case 11, Case 21, Case 22,
Ru release fraction | Ru release Ru release Ru release
of 2x10° fraction of 4 fraction of 2x10° fraction of 4
O-.2 0 0 0 ie)
2-5 0 0 0 0
5-12 -0690 1870 1) ie)
1.2-16 .0331 1204 1.1329 4.1184
1.6-2.1 0633 .2926 1.3564 6.2750
2.1-3.2 .0945 6264 2.3060 15.2788
3.2 -4.0 1963 1.8980 1.0609 10.2574
4.0-4.8 .2923 2.3100 1.4521 11.4777
48-56 3523 2.2660 1.7357 11.1653
5.6 - 8.1 0564 10.9909 .2699 52.6050
8.1-11.3 0058 69.2661 0019 22.7135
11.3 - 16.1 0 1.1027 i} .3599
16.1 - 20.9 0 0 0 i)
Total 1.16 89.06 9.32 134.25

The staff aiso performed sensitivity calculations to determine which isotope in the ruthenium group is
responsible for the increase in consequences when a significant ruthenium release is included in the
consequence calculations. Sensitivity calculations were performed with different ruthenium-group
isotopes included in the consequence calculations. The ruthenium-group isotopes remaining after a year
of radioactive decay are Co-58, Co-60, Ru-103, and Ru-106. These cases were run starting with the
Base Case. The results of these calculations are shown in Table 6. These results show that the
dominant isotope in the ruthenium group is Ru-106.


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where f is the logarithmic standard deviation of the fragility estimate and 1.044 is the
difference between the 10% non-exceedance probability (NEP) standard normal variable
(-1.282) and the 1% NEP standardized normal variable (-2.326). F, is tabulated below for
various fragility logarithmic standard deviation B values.

B | Median/CDFM Capacity | Fy=(Ciowcucire)
(Csor/ Cope)

03 2.01 137

0.4 2.54 1.52

0.5 3.20 1.69

0.6 4.04 1.87

For structures such as the spent fuel pool, 8 typically ranges from 0.3 to 0.5. Ref.
10 shows that over this range of B, the computed seismic risk is not very sensitive
to B. Therefore, I recommend using a midpoint value for B of 0.4.

Step 2: Determine hazard exceedance frequency H,q»,, that corresponds to Cy,
from the hazard curve.
Step 3: Determine seismic risk P; from:
Py = 0.5 Hioy 0)

Table 1 presents the Peak Spectral Acceleration PSA seismic hazard estimates from Ref.
8 and 9 (LLNL93 results) for the Vermont Yankee and Robinson sites. In order to accurately
estimate the seismic risk for a seismic HCLPF capacity Cucrpr of:

Cucter = 1.2g PSA = 1176 cm/sec? PSA (8)
associated with Screening Level 2 for the Vermont Yankee site by rigorous convolution, it is
necessary to extrapolate the Ref. 9 hazard estimates down to the 2x10" exceedance frequency.
Also, intermediate values in Table 1 have been obtained by interpolation.

Table 2 compares the seismic risk of spent fuel pool failure for these two sites as
estimated by the following three methods:

1. Ref. 1 simplified approach, i.e., Eqn. (3).

2. Ref. 10 simplified approach, i.e., Steps 1 through 3 above.

3. Rigorous convolution of the hazard and fragility estimates.

A2B-17


The fire ignition frequency for the SFP facility is therefore estimated to be
9.6E-4/yr + 1.6E-3/yr = 2.6E-3/yr. A fire frequency value of 3E-3/yr will be used in the analysis
to provide additional margin and to account for any uncertainties in equipment configuration.

4.2.1.4 Basic Event Probability

Basic Event Basic Event Probability
lE-FIRE 3.0E-3

4.2.2 Top Event CRA - Control Room Alarms
4.2.2.1 Event Description and Timing

This event represents fire detection system failure to alarm in the control room or operator
failure to respond to the alarm. The proper conditions for an alarm are assumed to exist within a
few minutes of fire initiation. Failure to respond could be because of operator error (failure to
respond), failure of the detectors, or loss of indication because of electrical faults. Success for
this event is defined as the operator recognizing the alarm and responding to the fire. Failure of
this event is assumed to lead to a fire damage state where there is a loss of the SFPC system
and a loss of the plant power supply system. This event is quantified by fault tree FIR-CRA and
includes hardware and human failures.

4.2.2.2 Relevant Assumptions

e The SFP area is equipped with fire detectors which are alarmed in the control room.
However, the area is not equipped with an automatic fire suppression system.

¢ _ Fire alarms will be activated in the control room within a few minutes of the initiation of a
fire.

. Regular maintenance and testing is performed on the fire detection system and on the
control room annunciators.

. Procedures are available to guide operator response to a fire, and plant operators are
trained in these procedures (NEI commitment no. 2).

4.2.2.3 Quantification

Human Error Probabilities

One human failure event is modeled for this event (basic event HEP-DIAG-ALARM). The

operator may fail to respond to a signal or indication in the control room. The source for this
error rate is THERP (Table 20-23).

Appendix 2A A2A-26


Table 3.1 Time to Bulk Boiling, and Boil-off Rates

Time after Decay power Total heat Time to bulk | Boil-off rate Level
discharge from last core | load (MW) boiling (hr) (gpm) decrease
(days) (MW) (ft/nr)'
2 16.4 18.4 5.6 130 1.0
10 8.6 10.6 9.8 74 0.6
30 5.5 7.5 14 52 0.42
60 3.8 5.8 18 41 0.33

90 3.0 5.0 21 35 0.28
180 1.9 3.9 27 27 0.22
365 1.1 3.1 33 22 0.18 = 0.2

Notes: (1) using typical pool sizes, it is estimated that for BWRs, we have 1040 ff°/ft depth, and for PWRs, we
have 957 ft?/ft depth. Assume = 1000 ft*/t depth for level decreases resulting from boil-off.

In an SFP, the depth of water above the fuel is typically 23 to 25 feet. Subtracting 3 feet to
account for shielding requirements, it is estimated that approximately 20 feet of water will have
to boil-off before the start of fuel uncovery. Therefore, using the above table, the available time
for operator actions for the loss of cooling type accidents is estimated as follows:

For one-year-old fuel, the total boiloff time available equals the time to bulk boiling plus the time
to boildown to 3 ft above the top of the fuel. Therefore, the total time available for operator
action is as follows:

Total Time  =33hr + (20 ft)/(0.2 ft/hr)
= 133 hours

It is assumed that the operator will not use alternate systems (e.g., firewater) until after bulk
boiling begins and the level drops to below the suction of the cooling system. It is assumed that
the suction of the cooling system is 2 ft below the nominal pool level. Therefore, if bulk boiling
begins at 33 hours, and the boil-off rate is 0.2 ft/hr, then the total boiloff time available to provide
make-up using the firewater system to prevent fuel uncovery is as follows:

2ft

133 hrs -(Time to Bulk Boiling + Time for Boil-off) = 133 - (33hrs+ Oaftlhr

) = 133- 43hrs = 90hrs
3.4 General Assumptions

This analysis is based on the assumption that the commitments for procedures and equipment
proposed by NEI in their November 12, 1999, letter to Richard J. Barrett (Ref. 4) are adopted.
These are reproduced below:

1. Cask drop analyses will be performed or single failure-proof cranes will be in use for
handling of heavy loads, (i.e., phase Il of NUREG-0612 (Ref. 13) will be implemented).

2. Procedures and training of personnel will be in place to ensure that onsite and offsite
resources can be brought to bear during an event.

Appendix 2A A2A-10


The results for the low density storage rack are given in Figure 2. As can be seen, criticality
cannot be entirely ruled out on the basis of geometry alone. Therefore, we examined the
conservatism implicit in the methodology and assessed whether there is enough margin to not
require any additional measures for criticality control. There are two main sources of
conservatism in the analyses; using fuel at the most reactive state allowed for the configuration
and not crediting soluble boron. By relaxing the assumption that all of the fuel is at its peak
expected reactivity, we have demonstrated by analyzing several sample storage configurations
that the rack eigenvalue can be reduced to approximately 0.998 (see Table 1). The storage
configurations analyzed included placing a most reactive bundle every second, fourth, sixth and
eighth storage cell (see Figure 3). The assemblies used between the most reactive assembly
were defined by burning the 5 w/o U,,3, enriched Westinghouse 15x15 assembly to 55
GWD/MTU which is a typical discharge burnup for an assembly of this type. This study did not
examine all possible configurations so this value should be taken as an estimate only.
However, the study does suggest that scattering the most reactive fuel throughout the pool
would substantially reduce the risk of a criticality accident. It is difficult to entirely relax the
assumption of no soluble boron in the pool, but its presence will allow time for recovery actions
during an event that breaches the SFP liner and compresses the rack but does not rapidly
drain the pool.

Although not all-inclusive because all fuel and rack types were not explicitly considered, the
physical controls that were identified are generically applicable. The fuel used in this study is a
Westinghouse 15x15 assembly enriched to 5 w/o U,,;with no burnable absorbers. The
Westinghouse 15x15 assembly has been shown by others (Ref. 7) to be the most reactive
PWR fuel type when compared to a large number of different types of PWR fuel. Furthermore,
the use of 5 w/o U,,. enriched fuel will bound all available fuel types because it represents the
maximum allowed enrichment for commercial nuclear fuel.

BWR Spent Fuel Storage Racks

In these analyses, we differentiated between high and low density BWR racks. The
conservatism inherent in the analyses must be considered (for BWR racks, the use of the most
reactive fuel allowed only) when considering the discussion of these results. The results of the
analyses of high density BWR racks are given in Figure 4. As can be seen, criticality is
prevented by design for the high density configurations. The poison sheets remain reasonably
intact following the postulated compression event. The poison sheeting (in this case Boraflex)
is sufficient to maintain subcriticality.

The results of the low density BWR rack analyses are shown in Figure 5. Here, as with the
PWR low density racks, criticality cannot be prevented by design. Once again we assessed
the impact of eliminating some of the conservatism in the analyses which in the case of BWR
storage is only related to the reactivity of the assembly. Analyses were performed placing a
most reactive assembly in every second, fourth, sixth and eighth storage cell. The assemblies
placed between the most reactive assemblies were defined by burning the 4.12 w/o enriched
General Electric (GE) 12 assembly to 50 GWd/MTU_ These analyses demonstrate that it is
possible to reduce the rack eigenvalue to approximately 1.009 (see Table 1). As previously
mentioned, this study did not include all possible configurations so this value should be taken
as an estimate only. Because BWR pools are not borated, there is no conservatism from the
assumption of no soluble boron.

Appendix 3 A3-7


With regard to protection against malevolent use of land-based vehicle, the staff continues to
believe there is no compelling justification for requiring a vehicle barrier as perimeter protection
at this time for ISFSIs. The staff will, however, continue to review the requirements to ensure
that a proper level of security is provided for existing casks, new cask designs, and other
changing technologies.

Comment #74: For Security, the integrated decommissioning rule should allow licensees to be
excused from 10 CFR 73.55 requirements upon a showing that the consequences of sabotage
can not exceed a defined dose to the public at the site boundary.

Response: The staff believes that 10 CFR 73.55 should be modified to a level commensurate
with the risk associated with safeguarding permanently shutdown plants, but not to a level less
than that provided for an ISFSI as described in 10 CFR 73.51.

Comment #75: The report concludes that there is no methodology currently available to
access probabilities of terrorist activity or behaviors which might culminate in attempted
sabotage of spent fuel. We disagree. For instance, Sandia National Laboratories, a key
contractor employed by the NRC on security matters, has applied a probabilistic approach to
security in decommissioning on the Maine Yankee docket. We encourage the staff to review
this report.

Response: The staff reviewed the information provided by the commenter. After review, the
staff maintains that there is currently not an acceptable methodology available to access the
probability of terrorist activity. The report in question, its identity verified through NEI, is

“A Vulnerability Analysis of a Proposed Security Plan for the Maine Yankee Power Plant,” dated
January 9, 1998. The purpose of this report was twofold: first, it presents the results of an
analysis of the effectiveness of the proposed physical security system in preventing or
mitigating an attempt by the design basis threat adversaries attempting radiological sabotage,
and second, it presents the results of a study to determine the need for a vehicle barrier
systems. This report does not predict the probability of terrorist activities or behaviors. The
Staff has read this report, and conducted an on-site inspection (report dated June 8, 1999) of its
technical findings and found them to be deficient. Further information on the inspection can be
found in the June 8 inspection report, Inspection Report #:50-289/99-06.

Comments #76, 77 and 78: (A) It’s conspicuously absent from your review of risk in this overall
subject, that the staff hasn't looked at the issue of sabotage and terrorism. (B) The draft report
omitted acts of sabotage and vandalism. Emergency evacuation plans should be prepared
with this consideration of terrorism. (C) It is suggested that NRC “err on the side of safety”
since terrorist acts can not be specifically addressed.

Response: The commenters are correct that security is identified, but not highlighted, in the
report. The report is a technical study to quantify the risks as it relates to the draining of a
decommissioning plant's spent fuel pool and the issue of a zirconium fire. It was not intended
to address security in any detail. The staff will be addressing specific security requirements for
decommissioning plants in its integrated rulemaking, which is an outgrowth of the technical
study. As with any rulemaking, there will be opportunities for the public to comment on the
security requirements the staff is recommending.

Appendix 6 AG-25


between refueling outages is on the order of a year, short-lived radionuclides in the SFP will
decay away between refueling outages. As a result, all of the short-lived radionuclides are in
the core at the start of the 11" refueling outage for Millstone 1. When Millstone 1 discharged
one-third of its core at the beginning of the 11" refueling outage, two-thirds of its short-lived
isotopes remained in the vessel. Therefore, use of 11 batches of fuel in Case 2 without the
remaining two-thirds of the core represents about a factor-of-three reduction in short-lived
radionuclides in the SFP from what was modeled in Case 1. As shown in Table A4-8, use of 11
batches of spent fuel without the remaining two-thirds of the core resulted in a factor-of-two
reduction in the prompt fatalities and no change in the societal dose and cancer fatalities. This
factor-of-two reduction in prompt fatalities is consistent with the factor-of-three reduction in the
inventories of the short-lived radionuclides when the remaining two-thirds of the core in the
vessel is not included in the consequence calculation.

Table A4-8 Mean consequences for Case 2

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 89 44,900 2,280
0-500 89 557,000 25,100

90 days 0-100 .78 44,500 2,250
0-500 78 554,000 25,000

1 year 0-100 53 43,400 2,180
0-500 53 567,000 25,500

The results of the next case, Case 3, are shown in Table A4-9. This case used a generic
population distribution of 100 persons/mile? (uniform). This was done to facilitate comparison of
the consequence results with the results of the analyses in NUREG/CR-4982 and
NUREG/CR-6451. Use of a uniform population density of 100 persons/mile? results in an
order-of-magnitude increase in prompt fatalities and relatively small changes in the societal
dose and cancer fatalities.

Appendix 4 A4-10


Fire engine: 100 - 250 gpm [depending on hose size: 1-%” (100 gpm) or 2-%”
(250 gpm)]

It is therefore assumed that, for the larger loss of coolant inventory accidents, make-up
through the makeup pumps is not feasible unless the source of inventory loss can be
isolated.

The operators perform walk-downs of the SFP area once per shift (8- to 12-hour shifts). A
different crew member is assumed for the next shift. It is also assumed that the SFP water
is clear and pool level is observable via a measuring stick in the pool that can alert
operators to level changes.

Requirements for fire detection and suppression may be reduced (when compared to
those for an operating plant) and it is assumed that automatic detection and suppression
capability may not be present.

All equipment, including external sources (fire department), are available and in good
working order.

The emergency diesel generators and support systems such as residual heat removal and
service water (that could provide SFP cooling or make-up before the plant being
decommissioned) have been removed from service.

The SFP cooling system, its support systems, and the electric driven fire protection pump
are fed off the same electrical bus.

Procedures exist to mitigate small leaks from the SFP or for loss of the SFP cooling
system.

The only significant technical specification applicable to SFPs is the requirement for
radiation monitors to be operable when fuel is being moved. There are no technical
specifications requirements for the cooling pumps, makeup pumps, firewater pumps, or
any of the support systems.

There are multiple sources of water for make-up via the firewater pumps or fire engine.

Generic industry data were used for initiating event frequencies for the loss of offsite
power, the loss of pool cooling, and the loss of coolant inventory.

Instrumentation that measures SFP temperature and level measures these parameters
directly.

For the purposes of timing, the transfer of the last fuel from the reactor to the SFP is
assumed to have occurred one year previously.

Appendix 2A A2A-12


Table A4-6 Cases Examined Using the MACCS2 Consequence Code

Case | Population Radionuclide Evacuation La/Ce Evacuation
Distribution Inventory Start Time Release Percentage
Fraction
Base | Surry 11 batches plus | 1.4 hours 1x10 99.5%
Case rest of last core | after release
begins
1 Surry 11 batches plus | 1.4 hours 1x10 95%
rest of last core | after release
begins
2 Surry 11 batches 1.4 hours 1x10 95%
after release
begins
3 100 11 batches 1.4 hours 41x10% 95%
people/mi? after release
begins
4 100 11 batches plus | 1.4 hours 1x10% 95%
people/mi? rest of last core | after release
begins
5 100 11 batches plus | 3 hours 1x10° 95%
people/mi? rest of last core | before
release
begins
6 100 11 batches plus | 3 hours 6x10* 95%
people/mi? rest of last core | before
release
begins
7 100 11 batches plus | 3 hours 1x10* 99.5%
people/mi? rest of last core | before
release
begins

The Base Case was intended to model the offsite consequences for a severe SFP accident for
a decommissioned reactor. To accomplish this, the Base Case used the Millstone 1 inventories
from NUREG/CR-4982 adjusted for reactor power and the rest of the last core as discussed
above. Accordingly, the Base Case used the Millstone 1 radionuclide inventories for the fuel
from the first 11 refueling outages (1649 assemblies) together with the rest of the last core (413
assemblies). Because the Millstone 1 core design has 580 assemblies, the amount of fuel
assumed to be in the SFP is equivalent to about 3.5 cores.

Other modeling in the Base Case, such as the population distribution, the evacuation
percentage of 99.5% of the population, and the meteorology, are from the NUREG-1150

Appendix 4

A4-8


Table A2e-4 Combined Variation in Intensity Along Length
and Across Width of Tornado Path”

Local True maximum tornado state
tornado
state FO FA F2 F3 F4 FS
CV-FO 1.0 0.641 0.380 0.283 0.298 0.286
CV-F1 0.359 0.471 0.433 0.358 0.333
CV-F2 0.149 0.220 0.209 0.195
CV-F3 0.064 0.104 0.116
CV-F4 0.031 0.054
CV-F5 0.016
(*) - Table 15b from NUREG/CR-2944
Table A2e-5 Exceedance Probability for Each F-scale
NUREGICR-2044 Exceedance probability (per year)
Region FO FA F2 F3 F4 F5
A 74x10 4.4x10°5 1.5x10°% 3.5x10°% 5.6x10°" 3.1x10%
B 5.6x10°S 3.3x10°° 1.1x10°% 2.5x10°% 3.7x10°" 2.1x10
c 2.9x10°S 1.5x10°° 4.1x10% 8.9x10°7 1.3x10°7 4.7x10°
D 3.6x10% 1.6x10°° 3.9x10°" 8.7x10° 1.6x10°° _
USA 3.5x10% 2.0x10% 6.1x10% 1.4x10%° 2.2x10°7 1.0x10°°

Appendix 2E

A2E-6



Gary M. Holahan
Table 3 Results based on NUREG-1465 (modified) Source Term

3

Cas Decay Mean Consequences?
e Time (Surry population, 95% evacuation)

Individual Risk of | Individual Risk Societal Early

Early Fatality of Cancer Dose (rem) Fatalities

(within 1 mile) Fatality (within (within 50 (within 10

10 miles) miles) miles)
79a 30 days 4.43x107 8.24x10? 2.37x10" 191
79b 90 days 4.19x10? 8.20x10? 2.25x10" 162
79c 1 year 3.46x107 8.49x10? 1.93x107 76.9
79d 2 years 2.57x107 8.42x107 1.69x107 19.2
79e 5 years 8.96x10% 7.08x107 1.45x107 1.34
79f 10 4.68x10* 6.39x107 1.34x107 .360
years
80a° | 30 days 2.01x10* 4.79x10° 1.35x107 5.38
80b° | 90 days 1.87x10° 4.77x10° 1.29x107 3.61
80c? 1 year 1.50x10° 4.33x10% 1.12x107 -951
80d? 2 years 1.12x10° 3.70x10° 9.93x10® -148
80e° 5 years 3.99x10* 2.93x10° 8.69x10° 0162
8of 10 2.05x10% 2.64x10° 8.13x10° 00601
years

"Accident frequencies approximately 10°*/year or less.
*Based on early evacuation.



EXECUTIVE SUMMARY.

This report documents a study of spent fuel pool (SFP) accident risk at decommissioning
nuclear power plants. The study was undertaken to support development of a risk-informed
technical basis for reviewing exemption requests and a regulatory framework for integrated
rulemaking.

The staff published a draft study in February 2000 for public comment and significant comments.
were received from the public and the Advisory Committee on Reactor Safeguards (ACRS). To
address these comments the staff did further analyses and also added sensitivity studies on
evacuation timing to assess the risk significance of relaxed offsite emergency preparedness
requirements during decommissioning. The staff based its sensitivity assessment on the
guidance in Regulatory Guide (RG) 1.174, “An Approach for Using Probabilistic Risk
Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.”
The staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet
the design and operational characteristics assumed in the risk analysis. These characteristics
are identified in the study as industry decommissioning commitments (IDCs) and staff
decommissioning assumptions (SDAs). Provisions for confirmation of these characteristics
would need to be an integral part of rulemaking.

The results of the study indicate that the risk at SFPs is low and well within the Commission’s
Quantitative Health Objectives (QHOs). The risk is low because of the very low likelihood of a
zirconium fire even though the consequences from a zirconium fire could be serious. The
results are shown in Figures ES-1 and ES-2. Because of the importance of seismic events in
the analysis, and the considerable uncertainty in seismic hazard estimates, the results are
presented for both the Lawrence Livermore National Laboratory (LLNL) and the Electric Power
Research Institute (EPRI) seismic hazard estimates. In addition, to address a concern raised by
the ACRS, the results also include a sensitivity to a large ruthenium and fuel fines release
fraction. As illustrated in the figures, the risk is well below the QHOs for both the individual risk
of early fatality and the individual risk of latent cancer fatality.

The study includes use of a pool performance guideline (PPG) as an indicator of low risk at
decommissioning facilities. The recommended PPG value for events leading to uncovery of the
spent fuel was based on similarities in the consequences from a SFP zirconium fire to the
consequences from a large early release event at an operating reactor. A value equal to

the large early release frequency (LERF) criterion (1x10° per year) was recommended for the
PPG. By maintaining the frequency of events leading to uncovery of the spent fuel at
decommissioning facilities below the PPG, the risk from zirconium fires will be low and
consistent with the guidance in RG 1.174 for allowing changes to the plant licensing basis that
slightly increase risk. With one exception (the H.B. Robinson site) all Central and Eastern sites
which implement the IDCs and SDAs would be expected to meet the PPG regardless of whether
LLNL or EPRI seismic hazard estimates are assumed. The Robinson site would satisfy the
PPG if the EPRI hazard estimate is applied but not if the LLNL hazard is used. Therefore,
Western sites and Robinson would need to be considered on a site-specific basis because of
important differences in seismically induced failure potential of the SFPs.


PWR Adiabatic vs. Air cooled
Fuel Burnup of 60 GWD/MTU

30
325 LZ

n
o

2 20 =
+ Air Cooled
—€ 15 Va

240 Adiabatic

5 a
a

0.164 7 2 3 4 5
shutdown time (years)

ie (hou

ti

heatu

Figure 1A-2. Comparison of Adiabatic and Air-Cooled Heatup Times

Reference

4 Hermann, et.al., “Technical Support for a Proposed Decay Heat Guide Using
SAS2H/ORIGENCS Data,” NUREG/CR-5625, September 1994.

Appendix 1A A1A-6


occupies a 30 x 30 x 30 foot cube within the large building). This model accounts for damage
from the aircraft (including, for example, being clipped by a wing). The estimated frequency
range for significant damage to the support systems is 1.0x10"° to 1.0x1 0* per year. The mean
value is estimated to be 7.0x10* per year. The second case estimates the value for the loss of
a support system (power supply, heat exchanger or makeup water supply). Based on the DOE
model including the glide path and the wing and skid area this case assumes a 10 x 10 x 10 foot
structure (i.e., the support systems are housed in a small building). The estimated frequency of
support system damage ranges from 1.1x10° to 1.1x10° per year, with the mean estimated to
be 7.3x107 per year. The third case uses the point model for this 10x10 structure, and the
estimated value range is 2.4x10" to 1.1x10* per year, with the mean estimated to be 7.4x107°
per year. Depending on the model used and the target structure size, the mean value for an
aircraft damaging a support system is 7x10” per year or less. This is not the estimated
frequency of fuel uncovery or a zirconium fire caused by damage to the support systems, since
the frequency estimate does not include recovery, either on site or off site. As an initiator of
failure of a support system leading to fuel uncovery and a zirconium fire, an aircraft crash is
bounded by other more probable events. Recovery of the support system will reduce the
likelihood of spent fuel uncovery.

Overall, the likelihood of significant spent fuel pool damage from aircraft crashes is bounded by
other more likely catastrophic spent fuel pool failure and loss of cooling modes.

3.5.3 Tornadoes and High Winds

The staff performed a risk evaluation of tornado threats to spent fuel pools (details are in
Appendix 2E). The staff assumed that very severe tornadoes (F4 to F5 tornadoes on the Fujita
scale) would be required to cause catastrophic damage to a PWR or BWR spent fuel pool.
These tornados have wind speeds that result in damage characterized as “devastating” or
“incredible.” The staff then looked at the frequency of such tornadoes and the conditional
probability that if such a tornado hit the site, it would seriously damage the spent fuel pool. To
do this the staff examined the frequency and intensity of tornadoes the continental United
States, using the methods described in NUREG/CR-2944 (Ref. 7). The frequency of an F4 to
F5 tornado is estimated to be 5.6x107 per year for the Central United States, with a U.S.
average value of 2.2x10” per year.

The staff then considered what level of damage an F4 or F5 tornado could do to a spent fuel
pool. Based on the buildings housing the spent fuel pools and the thickness of the spent fuel
pools themselves, the conditional probability of catastrophic failure given a tornado missile is
very low. Hence, the overall frequency of catastrophic poo! failure caused by a tornado is
extremely low (i.e., the calculated frequency of such an event is less than 1x10° per year).

It was assumed that an F2 to F5 tornado would be required to significantly damage SFP support
systems (e.g., power supply, cooling pumps, heat exchanger, or makeup water supply). These
tornados have wind speeds that result in damage characterized as “significant,” “severe,” or
“worse.” The frequency of an F2 to F5 tornado is estimated to be 1.5x1 0° per year for the
Central United States, with a U.S. average value of 6.1x10° per year. This is not the estimated
frequency of fuel uncovery or a zirconium fire caused by damage to the support systems, since
the frequency estimate does not include recovery, either on site or off site. As an initiator of

3-24


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4.5.3 Top Event CRA - Control Room Alarms

4.5.3.1 Event description and Timing

This top event represents the failure of the control room operators to respond to the initial loss of
inventory from the SFP. This top event is represented by fault tree LOI-CRA. Depending on the
leak size, the timings for the water level to drop below the level alarm set point (assumed 1 ft
below the normal level) would vary. It is estimated that water level would drop below the low-
level alarm set point in about 4 hours in the case of a small leak and in the case of a large leak,
it would take 1 to 2 hours. Failure to respond could be because of operator failure to respond
to an alarm, or loss of instrumentation system. Success for this event is defined as the
operators recognizing the alarm as indicating a loss of inventory.

4.5.3.2 Relevant Assumptions

Appendix 2A

A2A-54


equipment and tools matched to procedure, and crews that are. conversant with the procedures
and one another through training. were assumed .

HEP-FW-REP-DEPEN represents the failure of the repair crew to repair a firewater pump. Note
that the repair crew had failed to restore the SFPC system. Therefore, dependency was
modeled in the failure to repair firewater system. We assume that the operator will focus his
recovery efforts on only one pump. Assuming that it takes another two shifts (16 hours) before
technical help and parts arrive, then the operator has 72 hours (88 hours less 16 hours) to
repair the pump. Assuming a 10-hour mean time to repair, the probability of failure to repair the
pump would be Exp [-(1/10) * 72] = 1.0E-3. For HEP-FW-REP-DEPEN a low level of
dependence was applied modifying the nominal failure probability of 1.0E-3 to 5.0E-2 using the
THERP formulation for low dependence.

Functional fault tree LOC-OFD-L is similar except that basic event HEP-RECG-FWSTART is
replaced by HEP-RECG-FWSTART-L. The probability of this event is 5E-2, representing a low
level of dependence because of the fact that a failure to detect the condition during the first few
shifts may be indicative of a more serious underlying problem.

Hardware Failure Probabilities

Basic event FP-2PUMPS-FTF represents the failure of both firewater pumps. The pump may
be required to run 8 to 10 hours at the most (250 gpm capacity), given that the water inventory
drops by 20 ft (i.e., 3 ft from the top of the fuel). A failure probability of 3.7E-3 for failure to start
and run for the electric pump and 0.18 for the diesel driven pump are used from INEL-96/0334
(Ref. 14). Note that the relatively high unavailability assumed for the diesel driven firewater
pump may be conservative if it is subject to a maintenance and testing program, and there are
controls on availability. These individual pump failures result in a value of 6.7E-4 for event
FP-2PUMPS-FTF.

4.1.5.4 Basic Event Probabilities

Basic Event Basic Event Probability
HEP-RECG-FWSTART 2.0E-5
HEP-RECG-FWSTART-L 5.0E-2
HEP-FW-START 1.0E-5
HEP-FW-REP-DEPEN 5.0E-2
FEP-2PUMPS-FTF. 6.7E-4

4.1.6 Top Event OFB — Operator Recovery Using Offsite Sources

4.1.6.1 Event Description and Timing

This event accounts for recovery of coolant make-up using offsite sources given the failure of
recovery actions using onsite sources. Adequate time is available for this action, provided that
the operating staff recognizes that recovery of cooling using onsite sources will not be
successful, and that offsite sources are the only viable alternatives. This top event is quantified

Appendix 2A A2A-21


Table A4-9 Mean Consequences for Case 3

Decay Time in Distance (miles) | Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 11.7 50,100 2,440
0-500 11.7 449,000 20,300

90 days 0-100 10.6 50,300 2,460
0-500 10.6 447,000 20,200

1 year 0-100 8.19 49,000 2,380
0-500 8.19 453,000 20,500

The results of the next case, Case 4, are shown in Table A4-10. This case includes the
remaining two-thirds of the core in the vessel. This was done to facilitate comparison of the
consequence results with the results of the analysis in NUREG/CR-6451. As discussed above
in the comparison of Case 1 with Case 2, this increases the prompt fatalities by about a factor
of two with no change in the societal dose or cancer fatalities.

Table A4-10 Mean Consequences for Case 4

Decay Time in Distance (miles) Prompt Societal Dose Cancer Fatalities

Spent Fuel Pool Fatalities (person-Sv)

30 days 0-100 18.3 53,500 2,610
0-500 18.3 454,000 20,600

90 days 0-100 16.3 52,100 2,560
0-500 16.3 465,000 21,100

1 year 0-100 12.7 50,900 2,490
0-500 12.7 477,000 21,600

Heat up of fuel in an SFP following a complete loss of coolant takes much longer than in some
reactor accidents. Therefore, it may be possible to begin evacuating before the release begins.
Case 5, which uses an evacuation start time of three hours before the release begins, was
performed to assess the impact of early evacuation. As shown in Table A4-11, prompt fatalities
were significantly reduced and societal dose and cancer fatalities remained unchanged.

Appendix 4


the global behavior of the SFP. It is our recommendation that the modeling of SFP accidents be
performed with codes capable of calculating the heat generation and losses associated with the
range of accidents, including phenomena associated with both water boiloff and air circulation.
Further, the calculation of critical decay times for establishing both the validity of ad hoc
evacuation and precluding fission product release must also include consideration of the
exothermic energy of reactions (i.e., reactions with air and steam) with cladding, or alternatively
demonstrate that such energy contribution is negligible in comparison to decay heat at that
point. Severe accident codes, such as MELCOR, developed for modeling the degradation of
reactor cores, would seem to be a reasonable approach for analysis of integral behavior and
would possess the general capabilities for modeling liquid levels and vapor generation, air
circulation, cladding oxidation and fission product release. Use of a severe accident code also
facilitates the use of self consistent modeling and assumptions for the analysis. The proper
calculation of fission product releases depends in large part on the prediction of thermal-
hydraulic conditions. More detailed CFD modeling would improve the calculation of boundary
conditions for air circulation and could be used in conjunction with integral codes to better
evaluate convective cooling. The kinetics of cladding reactions should be confirmed with
experiments designed to simulate the range of conditions of interest under steady state and
transient heating. The experimental database on ruthenium releases under conditions
applicable to SFP accidents is inadequate and we are currently extrapolating data from
conditions which tend to maximize such releases.

While there is uncertainty in the analysis of spent fuel degradation, especially for the conditions
of air ingression, it is also true that elements of the analysis contain conservatism. The
assumption of 75-100 percent release of ruthenium initiated at lower temperatures is based in
large part on tests with bare fuel pellets, testing of cladded fuel indicates that the cladding acts
as a getter of oxygen limiting release of ruthenium until virtually all of the cladding has oxidized.
Further, before significant ruthenium release occurs (in its more volatile oxide form) the
surrounding fuel matrix must be oxidized. During transient heatup of an SFP with temperature
escalation one would expect the ruthenium release to follow the oxidation of the cladding at
which point the fuel would more likely resemble a debris bed (the seismic event may also
contribute in that regard) limiting the release fraction. The competition between formation of
hyperstoichiometric UO, and UO, may also limit the release fraction below that seen in the data.
The use of a temperature criterion of 600 °C to preclude significant fission product releases is
conservative in that it is based in large part on data that discounts the effect of cladding to limit
releases. The cladding failures at low temperatures will still allow substantial retention of fuel
fines and the presence of unoxidized zircalloy will prevent formation of volatile forms of
ruthenium. More prototypic experimental data on releases under these kinds of conditions may
reveal that the onset of significant releases, especially ruthenium, would not occur under SFP
accident conditions until fuel rod temperatures reached much higher temperatures associated
with complete oxidation of the cladding.

Use of the hottest fuel assemblies to predict global release of fission products from the entire
spent fuel inventory is a significant conservatism as well. Transient fuel damage testing
indicates that at the time of local temperature escalation not all of the rod bundle undergoes
rapid heating, cooler regions can avoid the oxidation transient. Prediction of the propagation of
the temperature escalation to the cooler regions of the pool needs to be carefully examined to
see if significant benefit can be gained, at a minimum it will lengthen the period of fission product

Appendix 1B A1B-7


liner. Therefore, the staff estimates that the potential initiating frequency for crushing is
approximately 1.2X10° per year (based upon 100 lifts per year). The criticality calculations in
Appendix 3 show that even if the low-density BWR assemblies were crushed by a transfer cask,
it is “highly unlikely” that a configuration would be produced that would result in a severe
reactivity event, such as a steam explosion that could damage and drain the spent fuel pool.
The staff judges the chances of such a criticality event to be well below 1 chance in 100 even if
the transfer cask drops directly onto the assemblies. This would put the significant criticality
likelihood well below 1X10° per year.

Deformation of the low-density BWR racks by the dropped transfer cask was shown to most
likely not result in any criticality events. However, if some mode of criticality was to be induced
by the dropped transfer cask, it would likely be a small return to power for a very localized
region, rather than the severe response discussed in the paragraph above. This type of event
would have essentially no offsite (or onsite) consequences since the heat of the reaction would
be removed by localized boiling in the pool, and water would shield the site operating staff. The
reaction could be terminated with relative ease by the addition of boron to the pool. Therefore,
the staff believes that qualitative (as well as some quantitative) assessment of scenario 1
demonstrates that it poses no significant risk to the public from SFP operation while the fuel
remains stored in the pool.

With respect to scenario 2 (the gradual degradation of the Boraflex absorber material in high-
density storage racks), there is currently insufficient data to quantify the likelihood of criticality
due to the degradation. However, the current programs in place at operating plants to assess
the condition of the Boraflex and take remedial action if necessary provide sufficient confidence
that pool reactivity requirements will be satisfied. In order to meet the RG 1.174 safety principle
of maintaining sufficient safety margins, the staff judges that continuation of such programs into
the decommissioning phase should be considered at all plants until all high-density racks are
removed from the SFP. As such, SDA #7 should be considered in future regulatory activities
associated with SFP requirements. This additional assumption is identified as SDA #7.

SDA #7 Licensees will maintain a program to provide surveillance and monitoring of Boraflex
in high-density spent fuel racks until such time as spent fuel is no longer stored in
these high-density racks.

Based upon the above conclusions and the staff decommissioning assumption, the staff
believes that qualitative risk insights demonstrate conclusively that SFP criticality poses no
meaningful risk to the public.

3.7 Consequences and Risks of SFP Accidents

This section assesses the consequences and risks associated with SFP accidents. The
consequences are assessed in Section 3.7.1. Results are provided for both early evacuation
and late evacuation cases” to address the impact of evacuation on consequences, and for two

‘2 Early evacuation is initiated and completed before the SFP release. Late evacuation
is not completed before release.

3-27