N Comm t ited to Nucl
Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC
AUG 3 0 2005
U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Prairie lsland Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60
Response to Request for Additional Information Regarding the "Relief Request to Implement Risk-Informed lnservice lnspection (ISI) Scheduling for the Fourth 10-Year lns~ection Interval for Prairie lsland Units 1 and 2"
Reference: Letter from Nuclear Management Company, LLC (NMC) to Nuclear Regulatory Commission (NRC), "Relief Request to Implement Risk- Informed lnservice lnspection (ISI) Scheduling for the Fourth 10-Year lnspection Interval for Prairie lsland Units 1 and 2" dated December 29,2004.
Prairie lsland submitted a Relief Request to implement Risk-Informed IS1 Scheduling for the Fourth 10-Year lnspection Interval in a letter dated December 29, 2004 (Reference). By electronic mail, dated June 3, 2005, the NRC requested additional information regarding the relief request. The enclosure to this letter contains the response to that request.
Summarv of Commitments
This letter contains no new commitments and no revisions to existing commitments.
Thomas J. Palmisano Site Vice President, Prairie lsland Nuclear Generating Plant Nuclear Management Company, LLC
Enclosure
cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC
171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
ENCLOSURE
Response to Request for Additional Information Regarding the "Relief Request to Implement Risk-Informed lnservice lnspection (ISI) Scheduling for the Fourth I O - Year lnspection Interval for Prairie Island Units 1 and 2"
Response to Request for Additional Information, 4 pages plus
List of Acronyms, 1 page Attachment I , 17 pages Attachment 2,4 pages
Page 1 of 4
Enclosure Response to Request for Additional Information
The NRC questions are in bold type face. The NMC responses are in plain type face.
Did you exclude Class 2 pipe or welds that are exempt from American Society of Mechanical Engineers (ASME) inspection requirements from the population of welds evaluated in your RI-IS1 program? Both Regulatory Guide 1 .I78 and EPRl TR-112657 simply discuss Class 2 welds and do not differentiate between welds exempted from ASME inspection requirements and welds not exempted from these requirements. If you did exclude these Class 2 pipe welds from your RI-IS1 program, please identify the guidance you relied upon to exclude welds from your RI-IS1 program scope based on them being exempt from ASME inspection requirements.
There are two areas wherein exemption is taken for Class 2 welds. IWC-1220 provides exemption from ASME Section XI entirely (meaning that these welds are not included in Section XI scope). Table IWC-2500-1 includes an exemption from NDE if the thickness of the associated piping < 318" for piping > NPS4 or 5 115" for piping 1 NPS2 and 5 NPS4, however these exempted welds must be included in the total population.
Per a phone conversation with the Staff, NMC understands that the question is dealing with the exemption cited under IWC-1220(a) specifically. IWC-1220 exempts components from the volumetric and surface examination requirement of IWC-2500. NMC did not include those Class 2 piping welds that are exempt under IWC-1220.
The reason for NOT including the piping welds under IWC-1220 is that under a normal IS1 Program meeting the requirements of ASME Section XI these welds would not require volumetric examination nor would these welds be included in the total population of which the 7.5% is taken. The Risk-Informed Inservice Inspection Program (RI-ISI) is an alternative to the ASME Section XI Code requirements. And as stated in the NRC SER for the EPRl Topical Report, TR- 112657 Rev. B-A, "The staff concludes that the proposed RI-IS1 program as described in EPRl TR-112657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10CFR50.55a for the proposed alternative to the piping IS1 requirements with regard to the number of locations, locations of inspections, and methods of inspection". Since the welds exempted by IWC-1220 would not have been classified as Category C- F-I of C-F-2, there are no Section XI non-destructive examination (NDE) requirements and therefore no alternative is specified in the RI-IS1 Program for these welds.
2. On page 5 of your submittal, you describe the Westinghouse Owners Group probabilistic risk assessment (PRA) Peer Certification Review that was performed on the 1999 update PRA model. Per Regulatory Guide 1.178 dated September 2003, please list all Level A and B "Facts and
Page 2 of 4
Enclosure Response to Request for Additional Information
Observations" from the review and how they have been addressed in the Revision 1.2 model. If some of the Level A and B "Facts and Observations" have not been addressed, please state why they are not expected to result in model changes that could significantly affect the overall results or conclusions of the RI-IS1 consequence evaluation.
All closed Level A and B "Facts and Observations" are listed in Attachment 1, including the manner in which they have been addressed.
Attachment 2 lists the open Level B "Facts and Observations." For each item, the status is provided and there is either a discussion of potential impacts on RI- IS1 consequence evaluation or a statement that future PRA model updates will be evaluated for impact.
3. The Unit 1 and Unit 2 Reactor Coolant System in Tables 5-1-1 and 5-1-2 identify welds in the examination category B-F. Please specify if the welds in this examination category are piping welds or reactor vessel welds since the 1989 Edition of the ASME Code, Section XI, identifies dissimilar metal welds in B-F examination category to either the piping or the vessel welds. It is noted also that the risk-informed inservice inspection program in accordance with EPRl TR-112657, Revision B-A is applicable to the examination category B-F for piping welds.
Based on the conference call held with the staff, the inclusion of Category B-F welds that are associated with the vessel should not be included. The conversation focused on the nozzle-to-safe end welds that contain Alloy 600 material that is highly susceptible to Primary Water Stress Corrosion Cracking (PWSCC).
The plant has the following breakdown concerning Category B-F welds:
There are six ltem Number B5.10 welds (Reactor Vessel Nozzle-to-Safe End Butt Welds) There are five ltem Number B5.40 welds (Pressurizer Nozzle-to-Safe End Butt Welds) There are four ltem Number B5.70 welds (Steam Generator Nozzle-to-Safe End Butt welds)
These are all Nozzle-to-Safe End Butt Welds that are associated with vessels. However, between the two units, there is only one weld that includes material considered susceptible to PWSCC. This weld is off of the bottom of the Unit 2 pressurizer. This weld was selected for examination.
The NRC Safety Evaluation for the EPRl TR-112657 states "The staff concludes that the inclusion of B-F welds in a RI-IS1 Program is a plant-specific issue and that individual licensees should determine the safety significance of B-F welds and perform the examinations commensurate with the associated risk."
Page 3 of 4
Enclosure Response to Request for Additional Information
Since the weld containing material susceptible to PWSCC has been selected for examination, NMC believes that the Safety Evaluation intent has been met.
Page 4 of 4
List of Acronyms
AF AFW AOP ATW S CCDP CCF CDF CLERP CM cvcs DG ECCS EF EOP EPRl ET F&O HEP HRA INEL INSTAI R IPE LER LOCA LOCL LOlA LOOP LOSP M AAP MFW MS-FLB MSlV NMC PlNGP PM PORV PRA RCP RCS RHR RI-IS1 SBO SG SGTR S I SLOCA T&H VAC WOG
Auxiliary feedwater Auxiliary feedwater Abnormal operating procedure Anticipated transient without scram conditional core damage probability Common cause frequency Core damage frequency Conditional large early release probability Corrective maintenance Chemical and volume control system Diesel generator Emergency core cooling system Error factor Emergency operating procedure Electric Power Research Institute Event tree Facts and observations Human error probability Human reliability analysis Idaho National Engineering and Environmental Laboratory Loss of instrument air Individual plant examination License event report Loss of coolant accident Loss of cooling water Loss of instrument air Loss of offsite power Loss of offsite power Modular accident analysis program Main feedwater Main steam / main feedwater line break Main steam isolation valve Nuclear Management Company Prairie Island Nuclear Generating Plant Preventive maintenance Power-operated relief valve Probabilistic risk assessment Reactor coolant pump Reactor coolant system Residual heat removal Risk Informed - Inservice Inspection Station blackout Steam generator Steam generator tube rupture Safety injection Small loss of coolant accident Thermal hydraulic Volts, alternating current Westinghouse Owners' Group
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
09s) FRO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Page 1
of 17
Impact on R
I IS1
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
Level of
Significance B
A
Observation
Several items w
ere identified relative to initiating event identification and grouping.
(I) The basis for excluding from
the model challenges to the
POR
Vs post reactor trip is not adequately explained. T
his affects the initiating event grouping for E
vents 2, 8, 10, 16, 18, 19. A
dditionally, the model does not appear to directly consider
the consequences of a stuck open POR
V (no actual transfer to
the Small L
OC
A E
T). T
hough the plant has not actually experienced a PO
RV
opening following a transient, this does
not provide a sufficient basis for concluding that POR
Vs w
ill not open for all initiators in this class. A
ppendix D w
riteup (D
.12) shows that the PO
RV
-related event frequency contribution is sm
all (4.17E-5) and encom
passed by the contributions from
other Small L
OC
As. H
owever,
the new
(Rev 2) L
OC
A frequency for S
2 is 6E-5, so S
tuck Open PO
RV
s are no longer sm
all contributors to this class.
(2) Random
RC
P seal failure (i.e., a random failure resulting in
RC
P seal leakage greater than normal m
akeup capability) was
not included in the IE frequency for sm
all LO
CA
. Such
potential random R
CP seal failures have been assessed at
frequency in range 1 E-3 to 5E-3 by various sources. T
his event has been neglected in the IE
selection. The updated PI PR
A
frequency for SI due to other than random
RC
P seal L
OC
A is
5E-3. T
his is comparable to frequency of random
RC
P seal L
OC
A, so the event should be considered.
(3) The T
2 initiator (without a stuck open P
OR
V) does not
appear to be an input into the transient event tree sequences.
The dual-unit L
OSP initiator frequency calculation in file
V.SM
D.96.005 (R
ecalculation of LO
SP Initiator) appears to be in error. T
he calculation divides LO
SP into PL
C (plant
centered), Weather (W
RL
) and Grid L
oss (GR
L) events, w
hich is correct. Prairie Island has had 2 dual unit L
OS
P events in it's
21 year history (as of 1996 when file w
as made). In calculating
the exposure time, the calc assum
es 42 plant years for PI,
Item
1
Status &
Resolution
CL
OSE
D -
The PR
A M
odel Revision 1.2 includes m
any significant changes to fix problem
s with the L
OC
A sizes and inputs
into the SLO
CA
tree. The L
OC
A sizes have been changed
to reflect industry standards. The SL
OC
A includes breaks
from 318 to 2 inches. T
he ML
OC
A includes breaks from
2 to 6 inches. T
he LL
OC
A includes breaks greater than 6
inches.
For the issue dealing with event of a PO
RV
lifting during a transient and failing to com
pletely reclose, a separate PO
RV
LO
CA
gate has been added under the SLO
CA
tree. T
he POR
V L
OC
A gate includes the scenario of a PO
RV
lifting during a norm
al transient and during a steam line
break. The norm
al transient captures all transients that can challenge a PO
RV
.
For the issue dealing with the random
RC
P seal LO
CA
, a separate initiating event has been added under the R
CP
SEA
L L
OC
A event tree, w
hich is transferred to the SL
OC
A tree. A
random seal L
OC
A initiating frequency
was determ
ined by reviewing N
UR
EG
ICR
-5750 data.
The third issue w
ith the T2 initiator com
es from the
proposed model and docum
entation (by a contractor). W
e are not using that inform
ation in the updated model. A
ll initiators used in the original m
odel (I-TR
I, I-TR
2, I-TR
3 and I-T
R4) are inputs into the transient event tree.
The issues presented in this F
&O
have been resolved and im
plemented in the R
ev 1.2 model update as described
above. (Same assum
ptions were used in the R
ev 2.0 m
odel.) C
LO
SED
- T
he LO
SP initiating event frequency was re-calculated
accounting for two dual-unit L
OS
P events over the history
of the plant. The L
OO
P frequency was calculated to be
7.5E-2lyr. T
his does not include Bayesian updating.
The new
calculated LO
OP frequency w
as incorporated into
F&O
IE-I>
sub- elem
ent
1E-4, sub-
element l3
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Page 2
of 17
Item
Level of
Significance
B
F&
O
1E-6, sub-
l6
Status &
Resolution
the Rev 1.2 m
odel. This change w
ill have a significant affect on the C
DF. H
owever, w
ith the addition of Off-site
Power R
ecovery in the model and other recom
mended
changes, the contribution that LO
OP m
akes to CD
F decreases in the new
model. (from
35% to approx. 24%
).
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (Sam
e assumptions used in the R
ev 2.0 model.)
CL
OSE
D -
The initiating event data referenced in this F&
O w
as not incorporated into the R
ev 1.2 (or Rev 2.0) m
odel.
In the Rev 1.2 m
odel, LO
OP frequency w
as calculated by dividing the num
ber of dual unit events (2 per unit) by the num
ber of comm
ercial operating years. The L
OO
P frequency w
as determined to be 7.5E-21yr. T
his does not include a B
ayesian update. This is a conservative
approach.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel (and Rev 2.0) update as
described above.
Observation
because it counts unit 1 and unit 2 separately (to be consistent w
ith the generic LO
SP data). The resulting B
ayesian updated dual-unit LO
SP frequency is 0.03 16. But if the units are
counted individually, then it must be considered that a dual unit
LO
SP at unit 2 affects unit 1, as opposed to the way it w
as calculated, w
hich effectively assumes unit 1 and unit 2 are tw
o different sites. T
herefore, the WR
L and G
RL frequencies m
ust be doubled because a dual unit L
OSP at unit 2 affects unit 1.
Alternatively, the PI site could be considered as a single unit
and there would be 2 failures in 20 site-years. T
his would be in
conflict the generic data and would require m
odification of the generic exposure tim
e.
Bayesian update w
as used for LO
SP frequency. The B
ayesian update algorithm
used is very sensitive to the error factor chosen for the generic data. T
he mean value for the generic
prior distribution for LO
SP was 0.01 8 1 w
ith an EF of 1.4. T
he plant specific data show
s that 2 LO
SP events have occurred in 25.7 site years (corresponding to a plant-specific point estim
ate of 0.0788lyr). H
owever, the updated m
ean calculated using the B
ayesian code and these values is .0187 - which hardly m
oves the prior m
ean at all. If the EF on the prior w
ere changed to 5, then the updated m
ean would be .044/yr, apparently m
ore reflective of the plant experience.
The review
ers believe that several calculational mistakes w
ere m
ade in this analysis.
I) the EF of the prior is calculated assum
ing that a chi-squared distribution represents the generic data, based on 43 events. T
his produces a very low E
F, since this process ignores the site to site variability.
2) the Bayesian update algorithm
used is sensitive to the choice of E
F.
3) if the EF on the prior actually w
as 1.4, then uncertainty bounds of prior and plant specific data w
ould not overlap and it could be said that the prior is not from
the same data base as the
plant specific.
The latest L
OSP report from
INE
L (N
UR
EG
ICR
-5496) provides a generic m
ean across the country of .05lyr. The PR
A
should be able to defend the derivation of a value significantly less than this.
Impact on R
I IS1
RI-IS1 consequence
analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
07s) FRO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 1
Page 3 of 17
Impact on
RI IS1
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island P
RA
m
odel used to perform
Level of
Significance B
B
Observation
This com
ment w
as generated by a review of the failure database
being developed for PRA
Rev 2.
The review
ers identified several concerns with the data
reduction for LO
SP. The L
OSP frequency as calculated by this
work is 0.0181. T
he LO
SP as calculated by INE
L in
NU
RE
GIC
R-5496 is 0.05. T
his discrepancy is large considering the im
portance of the event to the overall PRA
results. In addition:
1) More than 75%
of the events in the EPR
I database (EPR
I- T
R-106306) have been screened out as not being applicable.
The review
ers checked the screening assessments for several
events. In several cases the screening criteria seemed optim
istic and used the clause that "pow
er could have been restored if necessary", or "if this event happened at pow
er, OSP [offsite
power] w
ould have been restored. Other tim
es it was stated
that an error occurred at shutdown that could not occur at
power. T
he screening of events appears to have been too optim
istic about events at shutdown that w
ere assumed to not be
possible at power.
2) The data base screens out all but 56 events. H
owever, the
LO
SP frequency is calculated as 43 events12347 yrs. There is
no explanation of the difference between 56 events and 43
events.
3) The basis for the exposure tim
e of 2347 reactor-years is unclear. In the R
IF component database the accum
ulated operating tim
e is listed as 2546 licensed years, 2472 critical years and 2402 com
rnerical years. If there have been 2402 com
mercial years of operation, at an average availability factor
of 80%, there should be 1920 full pow
er years of operation, not 2347. T
he "2347 reactor years" used for the LO
SP calculation obviously includes the tim
e spent at shutdown. If all refueling
LO
SP events are removed from
the failure list, then the time
spent at shutdown should also be rem
oved from the exposure
time.
The review
ers did not find a discussion of dual unit initiators and subsequent station response, although at least one such initiator (dual-unit loss of offsite pow
er) is identified and an associated frequency is included am
ong the initiating events.
After the review
, Prairie Island PRA
personnel clarified that three potential dual-unit initiating events w
ere identified: Loss
of Offsite Pow
er, Loss of Instrum
ent Air, and L
oss of Cooling
Item
Status & R
esolution
CL
OSE
D -
The initiating event data referenced in this F&
O w
as not incorporated into the R
ev 1.2 model or the R
ev 2.0 model.
In the Rev 1.2 m
odel, LO
OP frequency w
as calculated by dividing the num
ber of dual unit events (2 per unit) by the num
ber of comm
ercial operating years. The L
OO
P frequency w
as determined to be 7.5E
-21yr. This does not
include a Bayesian update. T
his is a conservative approach.
The issues presented in this F&
O have been resolved and
appropriate changes were incorporated into the R
ev 1.2 m
odel (and Rev 2.0 m
odel) as described above.
CL
OSE
D -
A tw
o-unit model has been created w
hich captures the dual unit initiators in R
ev 2.0 model. T
he effects and impacts
that the dual unit initiators (I-LO
OP, I-IN
STA
IR, I-L
OC
L)
have on Unit 1 and U
nit 2 are included in the Tw
o-Unit
model.
Dependencies and success criteria are factored
into the initiating event system fault trees. T
he dual unit
F&O
1E-8, sub-
element
AS-6, sub-
element
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Pa
ge
4o
f 17
Level of
Significance
B
B
Observation
Water. O
f these, only loss of offsite power is m
odeled as a dual-unit event affecting unit 1 (i.e., an event for w
hich the status of the opposite unit is considered in the accident sequences w
ith respect to availability of opposite unit equipm
ent). The others are not so treated, because their
baseline CD
F contribution (when considered as single-unit
events) is relatively small.
Given the dependence of prim
ary and secondary pressure relief on instrum
ent air, the loss of instrument air event should be
discussed, and possibly modeled, independently of other
transient events. The prim
ary POR
Vs or possibly the
primarylsecondary safety valves m
ay lift to provide pressure relief in this scenario (loss of IA
). This m
ay be a unique enough plant response to w
arrant special treatment. In addition,
challenging these valves results in an increase in the S2 LO
CA
or steam
line break initiating event frequency.
The G
eneral Transient event tree (Figure 4.2 in the A
ccident Sequence notebook) show
s that if a consequential POR
V
LO
CA
occurs, a transfer is made to the S1 L
OC
A event tree.
The S I L
OC
A size range has been defined as 318" to - 1 "
(actually 718"). How
ever, the equivalent flow area for a
primary PO
RV
is expected to be larger than this, and should probably be considered in the S
2 LO
CA
category.
Additionally, the transfer for the M
SLB
scenario is not included in the R
ev. 1.1 model.
Item
7
F&
O
AS-8,
element
AS-11,
sub-
Status & R
esolution
initiator effects on the Unit 112 results can be found by
reviewing the PR
A Q
uantification notebooks.
CL
OSE
D -
During the R
evision 1.2 PRA
model update, an initiating
event fault tree was created for the L
oss of Instrument A
ir. T
he new initiating event fault tree provides a m
ore accurate calculation of the risk involved w
ith removing air
compressors from
service. In addition, a review
of past L
OIA
events at PI was perform
ed. The sequence of events
involved with a L
OIA
showed a slow
decrease in air pressure such that a reactor trip occurred w
ithout challenging the pressurizer PO
RV
s (LE
R 96-02-00) or the
operators had enough time to prevent a reactor trip
(February 1996 event). These tw
o events were initiated by
a failure of the air dryer exhaust purge valve to close follow
ing a dryer operation. This line has been m
odified per design change 96SA
O 1, w
hich installed an automatic
isolation valve in the exhaust lines of 121 and 122 air dryer. B
ased on the above discussion and the fact that there is a low
contribution of the LO
IA to overall C
DF
results - this issue can be considered closed.
In addition, during the Revision 1.2 m
odel update, credit w
as given for the pressurizer POR
V air accum
ulator and therefore the dependence of prim
ary pressure relief on instrum
ent air has decreased. C
LO
SED
- T
he PRA
Model R
evision 1.2 was changed significantly to
fix problems w
ith the LO
CA
sizes and inputs into the SL
OC
A tree. T
he LO
CA
sizes have been changed to reflect industry standards. T
he SLO
CA
includes breaks from
318 to 2 inches. The M
LO
CA
includes breaks from 2
to 6 inches. The L
LO
CA
includes breaks greater than 6 inches.
For the issue dealing with event of a PO
RV
lifting during a transient and failing to com
pletely reclose, a separate PO
RV
LO
CA
gate has been added under the SLO
CA
tree.
Impact on R
I IS1
RI-IS1 consequence
analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FAC
TS &
OB
SER
VA
TIO
NS (F
&O
's) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 1 P
age 5 of 17
Item
8 9 10
Observation
Consequential steam
generator tube rupture (i.e., SGT
R
resulting from a transient that causes a large pressure
differential across the steam generator tubes, such as steam
line rupture or inadvertently opened and stuck secondary side relief or safety valve) is not m
odeled in the accident sequences.
The possibility of this consequential event should be addressed
in the PRA
.
The success criteria for A
F are incomplete for Steam
Line
Break E
vents. Specifically, they do not include the requirement
to isolate flow to the faulted SG
.
These observations relate to the R
evision 2. Event T
ree N
otebook provided in the peer review package.
Docum
entation detail is limited in som
e areas, and should be expanded. A
ctually, some of these details already exist in the
previous layer of notebooks; it would be useful to capture this
F&
O
AS-12,
sub- elem
ent
AS-14,
sub- elem
ent l7
AS-15,
sub- elem
ent 3
Level of
Sig
nifican
ce
B
B
C (item
s 1- 5
) B
(items 6-
12)
Status & R
esolution
The PO
RV
LO
CA
gate includes the scenario of a POR
V
lifting during a normal transient and during a steam
line break.
The norm
al transient captures all transients that can challenge a PO
RV
.
The issues presented in this F
&O
have been resolved and im
plemented in the R
ev 1.2 model update as described
above. (The sam
e modeling w
as used in the Rev 2.0
model.)
CL
OSE
D -
The steam
generators at Prairie Island are designed such that the tubes can w
ithstand full system dp across the tubes
from the prim
ary or secondary sides without sustaining any
consequential tube ruptures. Because of this, the
consequential tube rupture event following a prim
ary or secondary depressurization w
as not modeled.
CL
OSE
D -
Changes have been incorporated into the R
ev 1.2 model to
account for the issue stated in this F&O
. The initiating
event for a Steam L
ine Break U
pstream of the M
SIV has
been added under the gate for the respective steam
generator. In addition, the initiating event for a Steam
Line B
reak Dow
nstream of the M
SIV and the failure of the
respective SG
MSIV
to close has been added under both steam
generator gates. Therefore, the steam
generator that has a steam
line break upstream of the M
SIV O
R has a
MS
IV that fails to close on a steam
line break downstream
of the M
SIV w
ill be failed. The A
FW
flow w
ill be isolated to the faulted SG
.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (T
he same m
odeling was used in the R
ev 2.0 m
odel.)
CL
OSE
D -
Although this finding is related to docum
entation that was
not incorporated into the current PR
A m
odel, the event tree notebook docum
entation was updated.
More details are
provided in the event tree notebooks on initiating event
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No im
pact.
This F
&O
has been resolved and incorporated into the
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 P
ag
e6
of 17
Level of
Significance Status &
Resolution
groupings, accident sequence progression, event tree structure, event tree headings, and event tree accident sequence analysis.
Impact on R
I IS1
Prairie Island PRA
m
odel used to perform
RI-IS1 consequence
analysis.
Observation
information in one E
T notebook to assure com
pleteness and consistency is obtained and m
aintained for the future updates.
Specific observations noted are as follows (som
e references are specifically to the SG
TR
event tree discussion, but may also be
applicable to other initiating events):
1. E
vent progress is not described in detail (ESD
s do not have m
uch more inform
ation content than ET
s; they do not m
ake up for the lack of detailed description of the event, nodes, operator actions, E
OPs involved, etc.).
2. T
op event descriptions are not detailed (SG isolation
appears to be consisting of MSIV
closure only. What
about operator actions, termination of A
FW flow
in to the faulted SG
etc).
3. T
op events with operator actions are not clearly
delineated and the dependence among top events is not
indicated.
4. R
eferences to EO
Ps are not complete (in w
hich EO
P(s) and by w
hat means does the operator identify and isolate
a faulted SG?)
5. T
here should be a one-to-one correspondence between the
items listed in section 4.10 and A
ppendix D. A
summ
ary table m
ay do it.
6. W
hy is there no SGT
R-W
branching when SG
TR
-ST1
fails in the SGT
R event tree (there is one in the E
SD) ?
7. G
ive guidance on what happens to sequences that branch
into other ET
s and end successfully there: for example
SGT
R has a transfer into A
TW
S and is successful; is it a success, or sim
ply truncated because it is low frequency?
What is the criteria for term
inating event tree to event tree looping?
8. M
S-FLB
events need to be discussed; they have an additional event tree node of "failure to isolate faulted SG
", which m
akes the event tree different from the
transient ET
. SBO
event tree needs to be discussed.
9. W
here are the "qualitatively assessed" items in E
SDs?
10. W
hat is the process that transfers the system success
criteria and operator action definition/success/dependence inform
ation from Section 4 and A
ppendix D to the system
analysts and H
RA
analysts? A couple of sum
mary tables
may be used to organize the "w
ork orders" generated for
item
F&
O
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Pag
e 7 of 17
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island P
RA
m
odel used to perform
Level of
Significance
A
A
Observation
the system and H
RA
analysts.
1 1. W
hat about stuck open pressurizer POR
V after a L
OSP
event? (maybe after a loss of M
FW event also?!)
Generic T
&H
analyses show that the PO
RV
s are challenged after a L
OSP event.
12. W
hat happens to the events with R
CS break flow
s that are less than m
akeup capacity; how long does the C
VC
S have to run; w
hat happens if CV
CS fails; W
hat is the underlying assum
ption in not modeling them
with an
event tree (small frequency?) ?
Tw
o steam generator tube rupture m
odeling items w
ere noted:
The dependency betw
een having a faulted SG follow
ing a SG
TR
with overfill and a stuck open relief valve and the top
gates for depressurization and AF are not considered in the
SGT
R developm
ent. The A
F top logic credits feed to both SGs.
Though acceptable for m
ost cases, if there is a stuck open relief valve on the ruptured generators, operators are directed to isolate that generator (including A
F). This reduces the ability to
depressurize with the 1 SG
and AF to the faulted generator
being isolated. In SG
TR
, the AFW
success criteria require AFW
to 1 of 2 SG.
Feeding of the ruptured SG is allow
ed (as directed by the E
OP's). T
he success path at function AFW
therefore allows
feeding of the bad SG. Subsequent event tree headings ask for
isolation of the ruptured generator. The fault tree developm
ent only asks about closing of the M
SIV on the ruptured generator.
In reality, if the good generator could not be fed, the ruptured generator could not be isolated. If the bad generator is being fed, the sequence needs to transfer on the failure path at "isolation" and go into E
CA
3.113.2. The fault bee logic for
"isolation" needs to include logic that "failure" to isolate the ruptured generator can be caused by failure of the good generator to be fed. If the ruptured generator is being fed, it w
ill not be isolated.
Tw
o items w
ere noted regarding derivation of success criteria for accum
ulators using MA
AP 3b calculations.
A M
AA
P calculation was used to determ
ine that accumulators
are only necessary for design-basis LO
CA
s. The M
AA
P PWR
A
pplication Guidelines specifically state that M
AA
P is not an appropriate code for use in analyzing rapid-depressurization events such as larger L
OC
As.
Item
11
l2
Status &
Resolution
CL
OSE
D -
The updated m
odel (Rev 1.2) has been m
odified to address this issue. T
he initiating event for Steam G
enerator Tube
Rupture has been added under the respective steam
generator gate and SG
POR
V gate. T
herefore, the fault tree logic w
as modified as to fail the ability to feed and
depressurize the ruptured SG.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above.
(The sam
e modeling w
as used in the Rev 2.0
model.)
CL
OSE
D -
The PR
A M
odel Revision 1.2 w
as changed significantly to fix problem
s with the L
OC
A sizes and inputs into the
SLO
CA
tree. T
he LO
CA
sizes have been changed to reflect industry standards. T
he SLO
CA
includes breaks from
318 to 2 inches. The M
LO
CA
includes breaks from 2
to 6 inches. The L
LO
CA
includes breaks greater than 6
F&
O
AS-18,
sub- elem
ent 10
TH
-lt
sub- elem
ent
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 1
Page 8
of If
Level of
Significance
B
B
Observation
No basis w
as found for not including accumulators in Sm
all L
OC
A event trees in cases w
hen high pressure injection fails. A
MA
AP calculation w
ithout accumulators w
as available, but this case show
ed core damage.
The tim
ing for switchover to recirculation in an analysis
proposed for PRA
Rev. 2 seem
s very conservative. First, it is assum
ed that containment spray initiates even for sm
all LO
CA
s, thereby reducing the tim
e to drain the RW
ST. Second, a
calculation assuming low
pressure injection is used for the tim
ing of both high- and low-pressure recirculation.
If high pressure recirculation is needed, R
CS pressure m
ust be above the shutoff head of the R
HR
pumps so that no low
pressure injection flow
has occurred, greatly increasing the time before
reciruclation is required. This could be im
portant because the lineup for high pressure recirculation is the only local critical step in the recirculation procedure.
This local step is the reason
that timing is so critical.
The L
OC
A break size definitions for the PIN
GP PR
A are based
on different criteria than those for most other PR
As. T
his w
ould be acceptable if the underlying analyses provided sufficient basis for the definitions, but it appeared that the available analyses do not adequately support the selected definitions.
The follow
ing is a comparison of the definitions and their bases,
with focus on the injection phase, as discerned from
the Event
Tree Success C
riteria notebook: PIN
GP PR
A S 1 (Sm
all LO
CA
category 1) = breaks that are too
large to be accomm
odated by the normal charging system
and too sm
all to provide adequate decay heat removal through the
Item
13
l4
F&
O
TH
-4, subelem
ent 4 T
H-9
9
sub- elem
ent
Status &
Resolution
inches.
In addition to this change, the accumulator is required in
the success criteria of the LL
OC
A injection phase (111
accumulator and 112 R
HR
pump).
One accum
ulator is failed due to a break in the R
CS cold leg.
The SL
OC
A and M
LO
CA
event trees were changed to
require accumulator injection w
ith the RH
R pum
p injection (111 accum
ulator and 112 RH
R pum
p). One
accumulator is failed due to a break in the R
CS cold leg.
The issues presented in this F
&O
have been resolved and im
plemented in the R
ev 1.2 model update as described
above. Same assum
ptions were used in the R
ev 2.0 model.
CL
OSE
D -
This F&
O relates to an analysis perform
ed by a contractor. T
his was a proposed analysis that is not used in the current
model and w
ill not be used in the updated model (R
ev. 1.2 or R
ev 2.0). T
he current timing for sw
itchover that is used for the new
SLO
CA
size was calculated using a plant-
specific MA
AP run. T
his run indicates that containment
spray does not actuate for a small L
OC
A.
CL
OSE
D -
Because of the m
any questions related to this issue, Prairie Island has changed the L
OC
A sizes in the R
ev 1.2 model
to the standardized definition of LO
CA
breaks. The new
break sizes are SL
OC
A (318 -
2 inches), ML
OC
A (2-6
inches) and LL
OC
A (>
6 inches).
MA
AP runs w
ere reviewed to support the success criteria
for the new break sizes.
In addition, the new L
LO
CA
m
odeling requires accumulator injection during short-term
injection, w
hich is included in the typical plant PRA
L
LO
CA
.
Impact on R
I IS1
RI-IS1 consequence
analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
09s) FRO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 P
age 9 of 17
Level of
Significance O
bservation
break; range defined as 318" to - 1" diameter breaks.
PING
P PRA
S2 (Small L
OC
A category 2) =
breaks that do not depressurize to w
ithin the low head injection system
capability but are w
ithin the capability of the high head injection system,
and that are sufficiently large to provide decay heat removal via
the break; range defined as - 1" to 5" diameter breaks.
TY
PICA
L PR
A Sm
all LO
CA
= breaks that are too large to be
accomm
odated by the normal charging system
and too small to
depressurize to the high head injection setpoint sufficiently rapidly to avoid the need for decay heat rem
oval; typically 318" to 2" diam
eter breaks.
PING
P Medium
LO
CA
= breaks that are sufficiently large to
depressurize to the shutoff head of the RH
R pum
ps but small
enough to be within the capability of the high head injection
system, w
ith decay heat removal via the break; range defined as
5" to 12" diameter breaks.
TY
PICA
L M
edium L
OC
A =
breaks that are sufficiently large to depressurize to the high head injection setpoint but for w
hich pressure rem
ains above the RH
R pum
p shutoff head, with decay
heat removal via the break; typically 2" to 6" diam
eter breaks.
PING
P Large L
OC
A =
breaks beyond the capability of the high head injection system
but which do not require accum
ulator injection, w
ith decay heat removal via the break and shutdow
n reactivity insertion via borated injection; range defined as 12" and greater but less than the design basis LO
CA
break size.
PING
P DB
A L
arge LO
CA
= break size for w
hich accumulator
injection is required in addition to low head injection; range
defined as the design basis break size.
TY
PICA
L L
arge LO
CA
= breaks that are sufficiently large to
depressurize to the RH
R pum
p shutoff head, with decay heat
removal via the break and shutdow
n reactivity insertion via borated injection; typically >
6" diameter breaks.
Am
ong the implications of the above are the follow
ing:
The PIN
GP PR
A S1 SL
OC
A plant response and m
odeling should be sim
ilar to the SLO
CA
response and modeling for
typical plant PRA
s.
The PIN
GP PR
A S2 SL
OC
A plant response and m
odeling should be sim
ilar to the ML
OC
A response and m
odeling for typical plant PR
As.
Item
-
F&
O
Status & R
esolution
The initiating frequencies for the new
LO
CA
sizes were
calculated from N
UR
EG
ICR
-5750.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. Sam
e assumptions w
ere used in the Rev 2.0 m
odel.
Impact on R
I IS1
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Page 10 of 17
Impact on R
I IS1
No Im
pact.
This F
&O
has been resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
No Im
pact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
Level of
Significance
B
B
Observation
The PIN
GP PR
A M
LO
CA
assumes that a single train of high
head injection can mitigate w
hat is equivalent to the low end of
the large LO
CA
size range for typical plants, for which high
head injection is normally not credited.
The PIN
GP PR
A L
LO
CA
(non-DB
A) plant response and
modeling differs from
the LL
OC
A response and m
odeling for typical plant PR
As in that it does not include a requirem
ent for accum
ulator injection; the LL
OC
A D
BA
plant response and m
odeling is equivalent to that for typical PRA
s.
The Success C
riteria notebook provides some perspective on
the rationale for what w
as done. How
ever, the guidance review
ed does not explicitly state the approach to be used for determ
ining the need for and types of thermal/hydraulic
calculations necessary to support the PRA
success criteria. Several instances have been noted (in other F&
Os) for w
hich detailed analyses have been required, and the M
AA
P code was
used without sufficient justification or check for applicability.
As described in the Safeguards V
entilation System N
otebook, room
cooling requirements have been addressed for the
equipment m
odeled in the PRA
. This notebook presents a
discussion, with references to engineering calcs, regarding the
need for cooling for each such room. H
owever, in som
e cases, it is not clear that the rationale provided for not m
odeling room
cooling is sufficient. For example, for the R
elay Room
, it is stated that analyses have show
n that it is necessary to maintain
the temperature below
120 deg F, but that room heatup analysis
showed that the tem
perature would reach 120 deg F at 11 hours.
Then the statem
ent is made that "T
his provides sufficient time
for the operator to perform the corrective actions per C
37.9 A
OP2." W
hile there may indeed be sufficient tim
e to perform
corrective actions, there is no guarantee that the actions will be
performed. Since the tem
perature exceeds the allowable
Item
15
16
Status &
Resolution
CL
OSE
D -
Although not explicitly stated in the calculation folders,
there was a m
ethodology for determining w
hen a MA
AP
case should be used in determining success criteria. Som
e of the criteria used in this determ
ination include:
1) If tim
ings were needed for im
portant operator actions.
2) T
he amount of tim
e it took to draindown tanks (i.e.
RW
ST)
3) T
o relax the USA
R success criteria for certain
accidents.
Although no guidance is w
ritten down on w
hen to apply the M
AA
P code, the use of the MA
AP code to support the
current model, does not present a questionable analysis or
inaccurate results. The results and conclusions from
the current m
odel are not significantly affected by this finding. C
LO
SED
- A
s part of the system notebook upgrade project, the
Safeguards Ventilation N
otebook has been revised to address issues related to crediting operator actions to restore room
cooling for the Control R
oom, R
elay Room
and B
attery Room
. A sensitivity study w
as performed for
each room to determ
ine the significant of modeling room
cooling for the specified room
s. The analysis show
ed that m
odeling the room cooling contributes very little to the
overall CD
F value and was of low
safety significance. The
documentation is m
ore clear and complete.
F&
O
TH
-13, sub- elem
ent
TH
-16, sub- elem
ent
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
09s) FRO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 P
age 11 of 17
Level of
Significance
B
B
B
Observation
equipment tem
perature well w
ithin the PRA
mission tim
e, there is a dependency on room
cooling for this room that should
either be modeled or m
ore carefully analyzed.
The fault tree m
odel, for large, medium
, and some sm
all S2
LO
CA
s, credits EC
CS flow
to the faulted loop. Unless therm
al- hydraulic analyses exist to provide a basis for this, it w
ould be expected that the injection path associated w
ith the faulted loop is unavailable, and only the rem
aining path would be available
for success. The success criterion should be 1 of 2 pum
ps to the single intact R
CS loop.
The corrective m
aintenance unavailability basic event for the 120V
AC
IP Inverters is modeled incorrectly in the Fault T
ree. A
s modeled, w
ith an inverter out of service, the fault tree still allow
s power to be supplied from
the alternate AC
source through the inverter to the instrum
ent panel. The sam
e comm
ent m
ay also apply to other inverter (and output breaker) failure m
odels in the PRA
.
As described in the Safeguards V
entilation System N
otebook, m
om cooling requirem
ents have been addressed for the equipm
ent modeled in the PR
A. T
his notebook presents a discussion, w
ith references to engineering calcs, regarding the need for cooling for each such room
. How
ever, in some cases,
it is not clear that the rationale provided for not modeling room
cooling is sufficient.
For example, for the R
elay Room
, it is stated that analyses have
Item
17
l9
F&
O
TH
-17, sub- elem
ent
SY-2, sub-
element
SY-7, sub-
element
Status & R
esolution
CL
OSE
D -
The R
ev 1.2 model includes the necessary logic to rem
ove the faulted loop as a possible flow
path during LO
CA
s. L
oop specific LO
CA
initiating events have been added to the m
odel, which w
ill fail the appropriate RC
S injection loop. T
his results in success criteria of 1 out of 2 pumps to
the single intact RC
S loop. In addition, the accumulator on
the faulted loop is also failed in the logic and is not available for injection.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (Sam
e assumptions w
ere used in the Rev 2.0
model.)
CL
OSE
D -
For the Rev 1.2 m
odel, the 120V A
C Instrum
ent Power
fault tree was changed so that the C
M event w
as moved
higher in the tree so that if it fails all power supplies that
feed the bus through the inverter. This change w
as perform
ed for the following:
1 1 (2 1) Inverter 12 (22) Inverter 13 (23) Inverter 14 (24) Inverter 17 (27) Inverter 18 (28) Inverter
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. Sam
e assumptions w
ere used in the Rev 2.0 m
odel.
CL
OSE
D -
AS part of the system
notebook upgrade project, the Safeguards V
entilation Notebook has been revised to
address issues related to crediting operator actions to restore room
cooling for the Control R
oom, R
elay Room
and B
attery Room
. A
sensitivity study was perform
ed for each room
to determine the significant of m
odeling room
cooling for the specified rooms. T
he analysis showed that
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 P
age 12 of 17
Impact on R
I IS1
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
Level of
Significance
B
Observation
shown that it is necessary to m
aintain the temperature below
120 deg F, but that room
heatup analysis showed that the
temperature w
ould reach 120 deg F at l l hours. Then the
statement is m
ade that "This provides sufficient tim
e for the operator to perform
the corrective actions per C37.9 A
OP2."
While there m
ay indeed be sufficient time to perform
corrective actions, there is no guarantee that the actions w
ill be performed.
Since the temperature exceeds the allow
able equipment
temperature w
ell within the PR
A m
ission time, there is a
dependency on room cooling for this room
that should either be m
odeled or more carefully analyzed.
As another exam
ple, for the rooms housing 120V
AC
Instrument
Power equipm
ent, there is no discussion of ventilation requirem
ents in the notebook. The equipm
ent survivability discussion notes that room
cooling is required, and that 4 hours are available follow
ing loss of ventilation to re-establish ventilation. H
owever, actions to open doors or re-establish
cooling are not modeled in the fault tree.
One editorial problem
also pertains to the ventilation modeling.
Assum
ption 5 in the SI system notebook states that room
cooling is not required for SI in injection m
ode, but the assum
ption does not address recirculation mode. T
he room
heatup calculation actually assumed sum
p recirculation mode,
and that should be noted in the notebook. T
he POR
V Fault T
ree for Feed & B
leed is applied in sequences involving initiators that w
ould cause containment isolation on
an S signal. The fault tree takes no credit for the PO
RV
accum
ulators to allow the PO
RV
s to be used after isolation of the air supply, and also takes no credit for operator action to re- establish air to the containm
ent. As a result, the m
odel assumes
failure of both POR
Vs w
hen air is isolated to containment.
As a result of the assum
ption that the POR
V accum
ulators are not sufficient for Feed and B
leed in scenarios involving an S signal, the m
odel appears to be overly pessimistic regarding
credit for feed & bleed.
FR.H
. I Step 1 1 provides direction to the operators to re-establish air to containm
ent, so consideration should be given to m
odeling this action, along with associated
valve failure probabilities.
Item
20
Status & R
esolution
modeling the room
cooling contributes very little to the overall C
DF
value and was of low
safety significance. The
documentation is m
ore clear and complete. A
s far as the SI pum
p room issue, the SI System
Notebook w
as also updated and the assum
ptions on room cooling are m
ore detailed and clear. R
oom cooling is not required for the SI
pump room
during injection or recirculation phase per Safety E
valuation 375.
CL
OSE
D -
The P
OR
V accum
ulator has been added to the model. T
his w
ill provide a source of air to the POR
Vs for Feed and
Bleed operation w
hen air is isolated to containment.
The
Rev 1.2 m
odel will take credit for the pressurizer PO
RV
accum
ulator if instrument air is not available. T
his is based on the follow
ing: A
) Procedures instruct the operators that the accum
ulators are available for operating the POR
V if
instrument air is not available.
B)
Operators are trained in the use of these procedures.
C)
The m
odel will conservatively assum
e a high failure probability for the accum
ulator (approximately 0.5)
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. T
he same assum
ptions were used in the R
ev 2.0 m
odel.
F&
O
SY-17,
sub- l3
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
FO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 P
age 13 of 17
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F
&O
has been resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
No Im
pact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
Level of Significance
B
B
B
Observation
The operating hours for the D
5 and D6 diesels w
ere not calculated correctly. In file V
.SMD
.95.007, the exposure time
for the planned maintenance (PM
) and corrective maintenance
(CM
) unvailablilites is stated as 175,344 hours. This is the sam
e exposure tim
e as for D1D
2, and appears to be the full 1 1 years of operation in the database. D
5 and D6 w
ere not installed until 1993. T
he exposure time the C
M and PM
for D5 and D
6 should be about 24,000 hr. T
his increases the PM
and CM
unavailabilities by a factor of 4.
(The exposure tim
e for fail to start and fail to run is calculated correctly.)
Notebook V
.SMN
.92.028 states that 4kv breakers are included in the fault tree m
odels but are not comm
on caused together because the the com
ponents supplied by the breakers already include any breaker com
mon cause failures that have occurred.
The com
ponent boundaries for all components fed by these
breakers (pumps, buses) should be consistent so that breaker
failure rates and CC
F rates can be consistently applied.
There are also no C
CF events for bus feeder breakers.
Most PR
As treat 4kv breakers separately from
served com
ponents, and include separate CC
F events for the im
portant sets of breakers.
In Rev 1, w
hen the plant specific data was 0 failures in T
exposure tim
e, the failure rate was calculated by assum
ing 0.5 failures in T
exposure time. T
his is mathem
atically equivalent to using a B
ayesian update with a Jeffrey's prior. T
here is no w
ay of knowing if this estim
ate is reasonable or not. A m
ore technically sound approach is to use a generic prior for B
ayesian update. In Rev2, the data developm
ent has changed to use 0.3 failures in the exposure tim
e. There is no basis for this
practice, expecially when the R
ev 2 data makes significant use
of Bayesian process.
Item
21
22
23
Status &
Resolution
CL
OSE
D -
For the Rev 1.2 m
odel the exposure times for D
S/D6 w
ere re-evaluated and new
unavailabilities were re-calculated
based on the new values. T
he exposure time for the PM
and C
M for D
5D
6 w
as 2 1864 hours.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (T
he same data w
as used in the Rev 2.0 m
odel.)
CL
OSE
D -
The N
RC
issued this same question during the initial
review of the IPE
. A specific R
equest For Information
question was issued by the N
RC
related to the omission of
the CC
F modeling of circuit breakers and electrical
switchgear. T
he PI PRA
group response follows:
"Com
mon cause failures of circuit breakers and sw
itchgear w
ere not explicitly modeled, but com
mon cause failures of
loads supplied through the breakers, such as pumps, valves
and other components that can be attributable to com
mon
cause mechanism
s were m
odeled. This im
plicitly captures circuit breaker com
mon cause failures that are associated
with these com
ponents. As w
ith circuit breakers, comm
on sw
itchgear (in terms of function and the effects of failures)
are implicitly analyzed w
ith other failures, such as em
ergency diesel generator comm
on cause failures."
The N
RC
approved the IPE, including this m
odeling assum
ption. C
LO
SED
- T
he approach using 0.3 failures in the exposure time w
as not incorporated into the R
ev 1.2 or Rev 2.0 m
odels.
If Bayesian updating process is used in future m
odel revisions, the recom
mendations from
this F&O
will be
incorporated.
F&O
DA
-39 sub- elem
ent
DA
-87 sub-
element lo
DA-I O, Sub- elem
ent l7
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O9s) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 1
Page 14 of 17
Level of
Significance B
B
A
Observation
The num
ber of plant specific failures for CV
CS pum
ps in Rev
2.0 seems high - about 60-80. T
here is no reason to use B
ayesian update techniques when there are such a large num
ber of plant specific failures. In fact, since the plant specific failure rate is relatively high com
pared to generic sources, it could likely be show
n that the PI CV
CS pum
ps are not in the same
population as generic pumps and a B
ayesian update process should not be used.
The H
RA
documentation indicates that operator interview
s w
ere conducted when determ
ining the execution time of
procedure steps, but the values used appear to be generic.
Further, a "generic" value of 45 minutes is identified as the
shortest time to core dam
age for any accident. This value is
then used in the screening analysis for several operator actions w
here the time to core dam
age is being estimated. T
here doesn't appear to be a basis for the 45 m
inute value. Furtherm
ore, it not clear that this value is applicable to the actions m
odeled.
Tw
o of the ten most im
portant operator actions, AB
US27R
ESY
and N
12 1 DR
YX
XY
(sorted by FV), are quantified using
screening values. This is contrary to the PIN
GP PR
A
groundrules and industry guidance.
Item
24
25
26
F&O
DA
-l I, sub- elem
ent
HR
-6, sub- elem
ent lo
HR
-7, sub- elem
ent
Status & R
esolution
CL
OSE
D -
The C
VC
S data in question w
as not incorporated into the R
ev 1.2 model or the R
ev 2.0 model.
The current failure rates for the C
VC
S pumps are based on
plant specific data without a B
ayesian update.
If a Bayesian Process w
ill be used to update the data inform
ation, the recomm
endations from this F&
O w
ill be considered. C
LO
SED
- T
he HE
P that were determ
ined by this method have been
re-calculated. A
new H
EP screening criteria w
as used. T
he majority of the H
EPs increase using this value
resulting in a more conservative approach.
This F&
O can
be considered closed out. The new
values have been incorporated into the R
ev 1.2 model and the R
ev 2.0 m
odel.
CL
OSE
D -
AB
US27R
ESY
was rem
oved from the m
odel, as this is an action that w
ould not be performed during accident
conditions. A recent plant m
odification was added to the
instrument air system
fault tree which caused the
importance of operator action N
12 1DR
YX
XY
to decrease such that its Fussel-V
esely is -lE-04 w
hich is well below
the N
MC
criteria for use of detailed human error m
odeling. T
hese modifications w
ere incorporated into rev 1.2 of the m
odel. Following these m
odifications and others, a new
screening was perform
ed which identified tw
o new
operator actions that were above the screening criteria and
were quantified w
ith screening values. An A
SEP analysis
was perform
ed on both of these events so that now there
are not any important operator actions that w
ere quantified w
ith screening values.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (Sam
e assumptions w
ere used in the Rev 2.0
model.)
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Page 150f 17
Item
27
28
29
Observation
Based on the operator action sensitivity study perform
ed, there are several scenarios involving m
ultiple human error events.
Some of the dependencies appear to have been recognized, but
it was not intuitively obvious how
they were factored into the
quantification of conditional HE
Ps (e.g., FDB
LD
OPA
TY
). Several scenarios involve m
ore than 4 HE
Ps, and this raises a question regarding how
the operator actions are being placed w
ithin the model. T
he product of some of these m
ultiple HE
P scenarios result in total crew
failure probabilities less than 1E-
06, which appears to be optim
istic.
The local actions in the sw
itchover to containment sum
p recirculation are m
odeled as 4 actions that are easy to recall. In actuality there are 13 distinct actions and only 4 are given as critical. N
o justification is given for the non-critical steps. Even
accepting that the other 9 actions are not critical, they would
certainly affect the operator's ability to remem
ber the steps. In general there doesn't appear to be any evidence for the non- criticality of tasks or that the added com
plexity they introduce has been considered.
This F&
O relates to both guidance and docum
entation sub- elem
ents of QU
.
A quantification notebook describing the follow
ing items needs
to be created:
how the one-top C
DF m
odel is constructed (guidance);
how any technical adjustm
ents are made to the top of the
FT
or in the systems below
(beyond what is docum
ented in the system
and event tree notebooks) to allow
F&O
HR
-11, sub- elem
ent 27
HR
-15, sub-
l7
QU
-l3 sub-
Level of
Significance A
B
B
Status & R
esolution
CL
OSE
D -
A new
rev 1.2 model has been created that has
incorporated many of the peer review
team com
ments.
Am
ong them is the explicit m
odeling within the one top
fault tree of the dependant operator actions. The m
odel was
solved by setting all of the operator actions to 1 .O. T
he top 100 accident sequences, w
hich contributed over 95% of the
core damage, w
ere analyzed for dependant actions. The
HE
Ps in these sequences were ordered as to w
hen they w
ould be performed in tim
e and new conditional H
EPs
were calculated using N
UR
EG
ICR
-1278. The new
conditional H
EPs w
ere then modeled in the one top fault
tree and the mutually exclusive file w
as used to remove
any illogical cutsets.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel update as described above. (T
he same assum
ptions were used in the R
ev 2.0 m
odel.) C
LO
SED
- T
he three operator actions in question (HR
EC
IRC
SMY
, H
RE
CIR
CX
XY
and RE
CIR
CX
XY
) which all involve
switchover to recirculation w
ere revised to incorporate the fact that the local operator m
ust perform all local actions
up to the point in which the critical actions required for
success are performed. T
he local operator now has a
procedure to perform these actions such that they do not
need to be performed from
mem
ory. The revised H
EPs
were incorporated in the updated rev 1.2 m
odel.
The issues presented in this F
&O
have been resolved and im
plemented in the R
ev 1.2 model update as described
above. (Same values w
ere used in the Rev 2.0 m
odel.) C
LO
SED
- A
Quantification N
otebook was created detailing the R
ev 1.2 and R
ev 2.0 PRA
model results.
The notebook
contains sufficient guidance for performing the process and
sufficient detail to document the inputs and outputs of the
process.
The issues presented in this F
&O
have been resolved and im
plemented in the R
ev 1.2 model (and R
ev 2.0 model)
update as described above.
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis.
No Im
pact.
This F
&O
has been resolved and incorporated into the Prairie Island P
RA
m
odel used to perform
RI-IS1 consequence
analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
09s) FRO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1 .
Page 16 of 17
Item
30
Level of
Significance
B
B
F&
O
QU
-3, sub- elem
ent 8
sub-
Status &
Resolution
CL
OSE
D -
For the Rev 1.2 m
odel, recovery of offsite power w
as credited for the L
OO
P sequences.
The issues presented in this F&
O have been resolved and
implem
ented in the Rev 1.2 m
odel (and rev 2.0 model)
update as described above.
CL
OSE
D -
An extensive review
of the Rev 1.2 and R
ev 2.0 model
results (top cutsets, dominant accident sequences, initiating
events review, im
portance measures, m
odel asymm
etries, operator actions) has been perform
ed and is documented in
the Quantification N
otebook.
As w
ith all the PRA
calculation folders, a senior PRA
person has review
ed the results.
Fleet PRA
procedures have also been developed and im
plemented w
hich address the PRA
model m
aintenance issues.
Observation
quantification;
any special logic introduced to model sequences (flags,
etc.);
supporting files (such as MU
TE
X, R
EC
OV
ER
Y, .B
E, .T
C,
etch sum
mary inputloutput files;
results summ
ary files and conclusions (See QU
-5 also);
computer run param
eters;
type of computer and operating system
, list and version of executable codes used;
limitations of the code;
references to supporting model notebooks (E
T, system
, H
RA
, data) etc.
Modifications perform
ed in the one-top fault tree, such as creation of the A
FW-T
fault tree from the full A
FW tree, m
ust be docum
ented either in the quantification or system notebooks.
The contribution of L
OO
P sequences that lead to loss of cooling w
ater and instrument air could be greatly reduced if credit could
be given to recovery of offsite power w
ithin the calculated time
to core uncovery of 5 hours.
PRA
group procedure 3.001 A requires evaluation of PR
A
results when the m
odel is updated, and documentation in
accordance with PR
A group procedure 1.002A
. The procedure
indicates that the evaluation must include a review
of top cutsets and basic event im
portance measures to ensure that
dominant contributors to risk are m
odeled accurately and that dependent operator actions are treated appropriately, w
ith focus on understanding and addressing risk significant issues that have resulted from
the latest requantification.
For a full PRA
update, consideration should also be given to review
ing more than just dom
inant contributors and top cutsets, depending on the extent of m
odeling change. For example, the
in-progress Rev 2 m
odel upgrade may produce results that w
ill
Impact on R
I IS1
No Im
pact.
This F&
O has been
resolved and incorporated into the Prairie Island PR
A
model used to perform
R
I-IS1 consequence analysis. N
o Impact.
This F&
O has been
resolved for the Prairie Island PR
A m
odel used to perform
RI-IS1
consequence analysis.
PR
AIR
IE ISL
AN
D C
LO
SED
FA
CT
S & O
BSE
RV
AT
ION
S (F&
O's) F
RO
M T
HE
W
EST
ING
HO
USE
OW
NE
RS G
RO
UP
(WO
G) P
EE
R R
EV
IEW
PR
OC
ESS
Attachm
ent 1
Page 17 of 17
Impact on R
I IS1
risk importance contributors, and overall C
DFIL
ER
F values.
Status & R
esolution L
evel of Significance
Observation
require a deeper review than an exam
ination of top cutsets, top
Item
- F&
O
PR
AIR
IE ISL
AN
D O
PE
N F
AC
TS &
OB
SER
VA
TIO
NS (F
&O
's) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 2
Item
Page 1
of 4
F&O
SY-4, sub-
element
DA
-5, sub-
DA
-6y sub-
Observation
The I20 V
AC
Model does not include failures of
the 120 VA
C Panel (bus faults). T
hese are norm
ally modeled in m
ost PRA
s.
The com
mon cause failure m
odeling was based
on methods and data in N
UR
EG
ICR
-4780. A
lthough the methods in this docum
ent are still valid, the C
CF factors (num
erical values) are based on plant experience and judgm
ent prior to 1988. N
UR
EG
ICR
-6268 (INE
L) is a m
ore current source of com
mon cause data and should
be used in the next update. There are several beta
factors in the current model that are 0.1 to 0.4 in
value. (RH
R, C
ontainment Sprays, Fan coolers).
In light of the more recent data in N
UR
EG
ICR
- 6268, these beta values are high and should be revised.
Plant specific data used to support PRA
Rev. 1
was collected for the IPE
in 1988. Generic
failure rates were used extensively in the IPE
. In 1995, an updated data collection w
as performed
for AFW
pumps, D
G's, A
ir compressors,
Cooling w
ater pumps, S
I pumps, and R
HR
Level of
Significance B
B
B
Status &
Resolution
OPE
N -
Due to the low
probability of the Instrument Panel fault,
this modeling error is not expected to have a significant
impact on the results.
A sensitivity analysis w
as performed to determ
ine the risk significance of including the Instrum
ent Panel fault in the PR
A m
odel. Appropriate basic events w
ere added to the 120 V
AC
panel logic (Panels 11 l(21 I), 112(212), 113(213), and 114(214)).
Results from
the Rev 2.0 m
odel showed no increase in
CD
F or LE
RF w
ith this modeling change.
The next revision to the m
odel will include failures of
the 120 VA
C Panel (bus faults).
OPE
N -
While it is true that N
UR
EG
ICR
-6268 and it's associated database represent a m
ore current database for the analysis of com
mon cause failures (C
CF), until a
plant specific analysis has been performed using this
database, it cannot be determined that the C
CF factors
that are used in the Rev 2.0 m
odel are too high. A
current version of the C
CF
database will be utilized to
analyze the CC
F factors during the continuing update
process.
We recognize the need to update the C
CF
numbers and
have a schedule and plan to update the data. How
ever, the data is applicable and can still be used.
A data update project has been started w
hich will
address this F&O
. O
PEN
- W
e recognize the need to update the plant specific data and have a schedule and plan to update the data. H
owever, the "old" data is applicable and can still be
used.
Impact on R
I IS1
No im
pact.
A sensitivity analysis w
as performed
to determine the im
pact of including this failure in the 120 V
AC
fault tree. T
he sensitivity study showed
that the CC
DP
and CL
ER
P values associated w
ith small, m
edium, and
large LO
CA
s did not change from
those provided in Prairie Island's R
I-IS1 submittal. T
he results of the sensitivity analysis determ
ined that the resolution of this F&
O has no
impact on the results or conclusions
of the Prairie Island RI-IS1
submittal.
In our opinion, data from
NU
RE
GIC
R-4780 is applicable and
can still be used.
It is our intent to update the CC
F
numbers using a m
ore current database as part of the data update project.
Any changes in the PR
A results due
to this modeling revision w
ill be evaluated to determ
ine the impact on
the RI-IS1 results as part of the
"living" aspect of the RI-IS1
program.
- It is our intent to update the plant specific data using m
ore current inform
ation as part of the data update project.
Any changes in the P
RA
results due
PR
AIR
IE ISL
AN
D O
PE
N F
AC
TS &
OB
SER
VA
TIO
NS (F
&O
's) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 2 Page 2 of 4
Impact on R
I IS1
to this modeling revision w
ill be evaluated to determ
ine the impact on
the RI-IS1 results as part of the
"living" aspect of the RI-IS1
program.
The H
RA
analysis update to meet
the new standards has been
completed and w
ill be incorporated into the next m
odel update.
Any changes in the PR
A results due
to this modeling revision w
ill be evaluated to determ
ine the impact on
the RI-IS1 results as part of the
"living" aspect of the RI-IS1
program.
Status &
Resolution
A data update project has been started w
hich will
address this F&O
.
OPE
N -
The m
ethodology used to calculate the pre-initiator H
uman E
rror Probability (HE
P) is adequate. How
ever, the PR
A group recognizes the need to use an im
proved m
ethodology to perform the calculation. T
he HE
P analysis needs to be updated to new
standards.
A H
uman R
eliability Analysis (H
RA
) update to meet the
new standards has been com
pleted and will be
incorporated into the next model revision.
Level of
Significance
B
Observation
pumps, w
hich were selected on the basis of risk-
significance to the PRA
results. A larger data
development effort is underw
ay for Rev 2, but
this still limits the plant specific data period to
1995.
The observed status of the use of plant-specific
data, given the above, is the following:
(a) 6 components in the R
ev. 1 PRA
have failure rates based on plant-specific data through 1995;
(b) a limited num
ber of other components in
Rev. 1 have failure rates based on plant-specific
data through 1988;
(c) most of the failure rates in R
ev. 1 are generic;
(d) after the Rev. 2 update, data w
ill only be current through 1995.
The review
ers believe the PRA
relies too heavily on plant data that is not sufficiently current w
ith the as-operated plant.
The equation used to quantify latent errors is not
intuitive, and appears to be incorrect. T
he equation presented in the HR
A notebook
suggests that there is a time period in w
hich a com
ponent can be considered available after corrective m
aintenance (CM
) but prior to retest (assum
ed to be 4 hours). Conversely, the
equation implies that no retest is perform
ed follow
ing preventive maintenance (PM
). This
most likely does not reflect m
aintenance practices. Furtherm
ore, the peer review
guidance suggests that latent errors may be
screened when a post m
aintenance test is perform
ed.
The sum
mation of the PM
, test (T), and random
failure (R
F) frequencies does not have any
Item
4
F&
O
HR
-4, sub- elem
ent 6
PR
AIR
IE ISL
AN
D O
PE
N F
AC
TS &
OB
SER
VA
TIO
NS (F
&09s) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P F
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 2 P
age 3 of 4
Impact on R
I IS1
No im
pact.
In our opinion, documenting and
evaluating a cross comparison
between sim
ilar plants is not expected to have a significant im
pact on the results or conclusion provided in the Prairie Island R
I-IS1 subm
ittal.
Status &
Resolution
OP
EN
- A
Quantification N
otebook was created detailing the
Rev 1.2 PR
A m
odel results. The notebook contains a
thorough evaluation of the quantification results including review
of top cutsets, dominant accident
sequences, initiating events, importance m
easures, m
odel asymm
etries, and operator actions.
How
ever, a comparison of our results to sim
ilar plants w
as not performed. A
s part of the Mitigating System
Perform
ance Index (MSPI) project, a W
OG
Com
parison report w
ill be completed on PW
Rs.
The significant
systems (Safety Injection, R
esidual Heat R
emoval,
Auxiliary Feedw
ater, Com
ponent Cooling, E
mergency
Diesel G
enerators, and Cooling W
ater) will be
compared.
Results from
the Westinghouse M
SPI Cross C
omparison
document related to Prairie Island w
ill be addressed as part of the M
SPI Project by Decem
ber 2005. Once this
is completed this F&
O w
ill be considered closed.
Level of
Significance
B
Observation
physical meaning, as the term
s appear to be m
utually exclusive. In addition, for components
only exposed to latent error on a refueling outage frequency, the approach m
entions that the operators w
ould most likely find a latent error
prior to startup. For these cases, a TI value of 4
is assumed w
hich is very similar to the C
M
cases. How
ever, in practice, at-power
surveillance test intervals are being substituted for T
I values applied to components exposed to
latent error only during refueling (e.g., C
TR
AIN
AX
XZ
, CV
HC
SI IXX
Z). L
astly, it seem
s that the refueling frequency value of 8.55E
-05hr is artificially reducing the HE
P in these cases.
The Peer R
eview supplem
ental guidance (draft subtier criteria) states that, for a category 3 classification for this sub-elem
ent, one must
fulfill the following:
"The accident sequence results by sequence,
sequence types, and total should be reviewed and
compared to sim
ilar plants to assure reasonableness and to identify any exceptions.
A detailed description of the T
op 10 to 100 accident cutsets should be provided because they are im
portant in ensuring that the model results
are well understood and that m
odeling assum
ption impacts are likew
ise well know
n.
Similarly, the dom
inant accident sequences or functional failure groups should also be discussed. T
hese functional failure groups should be based on a schem
e similar to that
identified by NE
I in NE
I 91-04, Appendix B
."
A sum
mary of top sequences by initiating event
was provided, as w
as a listing of risk-important
systems and operator actions. D
etailed descriptions of cutsets w
ere not provided, nor w
as a comparison of results to sim
ilar plants.
Item
5
F&O
QU
-5, sub-
PR
AIR
IE ISL
AN
D O
PE
N F
AC
TS &
OB
SER
VA
TIO
NS (F
&O
's) FR
OM
TH
E
WE
STIN
GH
OU
SE O
WN
ER
S GR
OU
P (W
OG
) PE
ER
RE
VIE
W P
RO
CE
SS
Attachm
ent 2 Page 4 of 4
Item
Level of
Significance
B
F&
O
QU
-6, sub- 27
Observation
Neither a quantitative uncertainty analysis nor a
qualitative evaluation of significant sources of uncertainty are addressed.
Status & R
esolution
OPE
N -
A data update project has been started w
hich will
address this F&O
.
Impact on R
I IS1
No Im
pact.
In our opinion, the RI-IS1
application is unaffected by the results from
an uncertainty analysis since the R
I-IS1 program is based on
the results from propagating point
estimates through the m
odel.