n prairie lsland nuclear generating plant comm to nucl operated … · 2012. 11. 20. · 5...

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


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


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


A


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>

subelem

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.



implem


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


R


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


A


R


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


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.



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


A


R


No Im

pact.

This F&

O has been


A


R


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,

subelem

ent

AS-14,

subelem

ent l7

AS-15,

subelem

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

.


&O


plemented in the R


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

.



implem


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


RI-IS1

consequence analysis.

No Im

pact.

This F

&O


A


R


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


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


A


R


No Im

pact.

This F&

O has been

resolved and incorporated into the Prairie Island P

RA

m


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.



implem


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,

subelem

ent 10

TH

-lt

subelem

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

subelem

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.


&O


plemented in the R


above. Same assum


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


A


R


No Im

pact.

This F&

O has been


A


R


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.



implem


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


RI-IS1


No Im

pact.

This F&

O has been


A m


RI-IS1


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, subelem

ent

TH

-16, subelem

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, subelem

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.



implem



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



implem


odel update as described above. Sam

e assumptions w


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


A


R


No Im

pact.

This F&

O has been


A


R


No Im

pact.

This F&

O has been


A m


RI-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 P

age 12 of 17

Impact on R

I IS1

No Im

pact.

This F

&O


A


R


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)



implem


odel update as described above. T

he same assum


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


A


R


No Im

pact.

This F

&O

has been resolved for the Prairie Island PR

A m


RI-IS1


No Im

pact.

This F&

O has been


A m


RI-IS1


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.



implem



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


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 subelem

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, subelem

ent

HR

-6, subelem

ent lo

HR

-7, subelem

ent

Status & R

esolution

CL

OSE

D -

The C

VC

S data in question w


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.



implem



e assumptions w

ere used in the Rev 2.0

model.)

Impact on R

I IS1

No Im

pact.

This F&

O has been


A m


RI-IS1


No Im

pact.

This F&

O has been


A


R


No Im

pact.

This F&

O has been


A


R


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 subelem

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, subelem

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.



implem



he same assum


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.


&O


plemented in the R


above. (Same values w


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.


&O


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


A


R


No Im

pact.

This F

&O


A


R


No Im

pact.

This F

&O

has been resolved and incorporated into the Prairie Island P

RA

m


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, subelem

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.



implem


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


A


R

I-IS1 consequence analysis. N

o Impact.

This F&

O has been


A m


RI-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 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



ine the impact on



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



ine the impact on



program.

Status &

Resolution


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, subelem

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 -


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.

n prairie lsland nuclear generating plant comm to nucl operated … · 2012. 11. 20. · 5...

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