ML20149L208: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
 
Line 1: Line 1:
#REDIRECT [[IR 05000285/1997009]]
{{Adams
| number = ML20149L208
| issue date = 07/25/1997
| title = Summarizes 970721 Predecisional Enforcement Conference in Arlington,Tx Re Apparent Violation Identified in Insp Rept 50-285/97-09.Licensee Presented Summary of Causes & Corrective Actions.Attendance List & NRC Handout Encl
| author name = Howell A
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name = Gambhir S
| addressee affiliation = OMAHA PUBLIC POWER DISTRICT
| docket = 05000285
| license number =
| contact person =
| document report number = 50-285-97-09, 50-285-97-9, EA-97-280, NUDOCS 9707310192
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 111
}}
See also: [[see also::IR 05000285/1997009]]
 
=Text=
{{#Wiki_filter:.
.
..
._
.
'
i
e'
4
} 'S
p" "' 4 t
UNITED ST ATES
y
l
#$
%';
NUCLEAR REGULATORY COMMISSION
-
: .'' . . ~ , ,
~' j i
REGloN iv
i
!
N.
'
C
e
i
S
#*
f
611 RY AN PL AZ A ORIVE. SUIT E 400
j
o....*T
t
~
AR LINGTON, TEXAS 76011 8064
-
l
,
;
July 25, 1997
i
1
EA No. 97-280
l
S. K. Gambhir, Division Manager
Production Engineering
Omaha Public Power District
Fort Calhoun Station FC-2 4 Adm.
P.O. Box 399
Hwy. 75 - North of Fort Calhoun
Fort Calhoun, Nebraska 68023-0399
Dear Mr. Gambhir:
SUBJECT:
PREDECISIONAL ENFORCEMENT CONFERENCE SUMMARY
On July 21,1997 representatives of Omaha Public Power District met with NRC personnel
in the Region IV office located in Arlington, Texas to discuss the apparent violation
identified 'a NRC Inspection Report Number 50-285/97-09. The conference was held at
the request of Region IV.-
The licensee presented a summary of the causes for the apparent violation and their
corrective actions.
The attendance list, NRC handout, and the licensee's presentation are encloseo to this
summary. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of -
this summary and its enclosures will be placed in the NRC Public Document Room.
Sincerely,
b
h Arth r T. Howell 111, Director
Division of Reactor Safety
,
i
Enclosures:
1. Attendance List
2. Licensee Presentation
(
k
3. NRC Handout
i
01OObb
Docket No.: 50-285
License No.: DPR-40
I b! !! h.h, !
9707310192 970725
**
gDR
ADOCK 05000285
PDR
~
.
J
 
.
$
.
Omaha Public Power District
-2-
cc w/ enclosures:
James W. Tills, Manager
Nuclear Licensing
Omaha Public Power District
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 399
Hwy. 75 - North of Fort Calhoun
Fort Calhoun, Nebraska 68023-0399
James W. Chase, Manager
Fort Calhoun Station
P.O. Box 399
Fort Calhoun, Nebraska 68023
Perry D. Robinson, Esq.
Winston & Strawn
1400 L. Street, N.W.
Washington, D.C. 20005-3502
Chairman
Washington County Board of Supervisors
Blair, Nebraska 68008
Cheryl Rogers, LLRW Program Manager
Environmental Protection Section
Nebraska Department of Health
301 Centennial Mall, South
P.O. Box 95007
Lincoln, Nebraska 68509-5007
 
-
.
Omaha Public Power District
-3-
l
E-Mail report to T. Boyce (THB)
E-Mail report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK)
bec to DCD (IE01).
bec distrib. by RIV:
'
Regional Administrator
DRS-PSB
DRP Director
MIS Systern
Branch Chief (DRP/B)
RIV File
Project Engineer (DRP/B)
Branch Chief (DRP/TSS)
Resident inspector
;
DOCUMENT NAME: R:\\_FC\\FCSUM.JL
To receive copy of document, indicate in box: "C"
Copy without enclosures
"E" = Copy with enclosures "N" = No copy
SRA:DRS
,l E
D:DRPA g
D:DRS [r),Y/, E
l
l
JFShackelfordJ
TPG W M
ATHo % I)I
s
07d_f97
l,
074fS7
1
07/25/97 "
OFFICIAL RECORD COPY
 
n
Omaha Public Power District
-3-
,
.4
E-Mail report to T Boyce (THB)
E-Mail report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK)
bec to DCD W
. bec distrib. by RIV:
~ Regional Administrator
DRS-PSB
DRP Director
MIS System
Branch Chief (DRP/B)
RIV File
Project Engineer (DRP/B)
Branch Chief (DRP/TSS)
Resident inspector
.
DOCUMENT NAME: R:\\_FC\\FCSUM.JL
To receive copy of document, indicate in box: "C"
Copy without enclosures "E" = Copy with enclosures "N" = No copy
SRA:DRS
E
D:DRPA, ,gg D:DRS f,Y/, E
,
JFShackelfor4
TPG W M
ATHoMMI
07F/97
[,
074Ff97
1
07/25/97 "
OFFICIAL RECORD CO')Y
 
.
-
-
-
._ _
.
.-
.
..
._
. . .
1
J
, .
'
,
I
.
I
ENCLOSURE 1
1
Fort Calb.aun Station
,
'
Predecisional Enforcement Conference Attendance List
!
l
l
l
i
 
._
_ - - _ _ _
_
-
_
__
.
. _ _ _ _ _ . _ _
_ _ . _
'
.
'
'
PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE
j
LICENSEE / FACILITY
Omaha Public Power District
Fort Calhoun Station
DATE/ TIME
July 21. 1997. 10:30 a.m.
CONFERENCE LOCATION
Region IV. Training Conference Room
.
Arlington. TX
{
EA NUMBER
EA 97-280
NRC REPRESENTATIVES
,
)
NAME (PLEASE PRINT)
ORGANIZATION
TITLE
;
W l l? H 1 C M (4 fn 6 L f L % H
HK&
0%V1y
ofpic sa
o
Fut3
mMscHoW
H%
RMIoans
(40knmqmfoe
b
94WP2foEd)
/U M
9 . M M 7tt h
sch CGJ NL
%k Owas
uec
A4k4d' Ihssw2
Nac 2iV
G4~
s %.u tE
~/
v
DAlf A. Ecujweg
NRc.
RS
CHtEF, /1&idre4gcc ?gM::s
*
/f.
. h/fGW
&C
RT
/ ES/GA/A L {cc&dfEL
V' Ni6 L
NRC/N2R
D(of
6.k@ W
l
Pd.t h
hW
ec m
'
W :llta , O. SL s. ,
W R c- Ki]L
R n <i ct,i e-f/DRP
Page _ of _
 
.-
-
.
.-
.
-.- .
. - . . -.
-.
-
"
'
.
,
!-
PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE
,
l
LICENSEE / FACILITY
Omaha Public Power District
'
Fort Calhoun Station
i
DATE/ TIME
July 21, 1997. 10:30 a.m.
!
CONFERENCE LOCATION
Region IV. Training Conference Room
Arlington. TX
EA NUMBER
EA 97-280
s
;
LICENSEE REPRESENTATIVES
L
NAME (PLEASE PRINT)
ORGANIZATION
TITLE
%nta
c 4re<
c> f90
vp-acu su
'
g u g ,., e x. 6 a n n u
oppo
Di'''^^ M " P "
E g in"~'1
'
04 o}M nD~3
M orl
l
&& X/dae, war
G46,0
m aa+ w cewe m s
1
;
Jemes W Y;7h
0/Yb
% a f a - t % d u , G a d ,9
i
don ld. u e m
6fPC
%
% ROW NY-
y ,,,e
Hwchu hou
l
Wenn<u s.Dow A
oPPD
p) _ . mn l. gLb.
"
doseph KGaspe<
om
*"M*"
.
l
k,b, kWCKGWJ
ONQ
Atacerette Asussinenb
/
VW/J/C*V N/J W
-
svue seaues aose
DAU10 l .
OLL /Wh
D hh0
' 0'd6Mm Even - EMiut/tonoved
l
,
Page _ of _
 
1
*
l
.
ENCLOSURE 2
;
Fort Calhoun Station
Licensee Presentation
l
 
;j-
.
.
,
!
l
OMAHA
!
PUBLIC POWER .
!lo
DISTRICT
i
:
i
!
Fort Calhoun Station
;
Main':enance Rule
i
!
Prec ecisional Enforcement Conference
4
f
l
July 21,1997
:
!
!
!
)
llO
:
l
1
l
 
_
_
_ _
_ _
_ _-
_______
e
:
-
e
... ... ;.
.
-A.ng.
*1ggg g
-
--
,
Opening Remarks
Introductions
e
Gary Gates
e
2
 
- _
- - .
-
- - -
-
- . - - - - -
- -
.
.
$
l
l
l
.
<
lO
;
;
4
J
-
. , ,,, n
.
. . .
. , , , . . . . . . . . . _ . _ _ . . , . .
..
-
..
I
!
l
:
.i
4
l
1
i
)
.
!
'
4
;
l
Agenda
~
,
f
-
~
O
Sudesh Gam ahir
O
3
._.
 
-.
>
.
_
Agenda
;
- _ .
o Operational Overview
Ross Ric enoure
o Maintenance Rule
Jim Tills / John Johnson / Ken Dowdy
-
o Assessments and Corrective Action
O
Update
Joe Gas aer
o OE Program Assessment
Dick Andrews
o Summary and OPPD Persaective
Sudesh Gambhir
o Closing Remarks
O
Gary Gates
4
 
_
_ _
_ _
- _ _ - - _ _ _
e
1
-
i
,h
.
.
i o
,
ii
TYfMd7W 'FV.]
#
;
.
.,
i
_ _ . . . . .. . . , , . ..
!
!
;
!
I
i
)
!
l
l
Operational Overview
,
;O
,
a
$
!
Ross Ridenoure
:
i
l'
1
l
'
,
!
1
l
i
!
l
!
i
!!O
:
1
i
'
1
5
i
!
- -
-
. - - .
 
-_
_ _
. - _ _ - _ _ - - - _ _ - - . - - - . -
*
)
.
;
Operational Overview
.
.
O
-
.
1
.,-sesvp pg
tg,g. gy p ,, _
, -
~
o Event Overview.
o Primary Plant Impact.
;
.
o Secondary Plant Imaact.
O
o Operational Safety Significance.
.
O
6
-
 
________-_-_
_-
,
*
,
l Overview of the Rupture.
!O
i
.
l
xv;n~;._
l
i o Major steam rupture in 4th stage
I
extraction steam line.
;
i
i
l o Reactor tripped within 19 seconds.
:
! O Crew entered emergency procedures
and stabilized the plant quickly.
o Crew responded in a decisive, timely
manner.
o NOUE declared and ERO activated.
O
7
 
l
Primary Plant Impact
-
!
e
:
-
.. i f " *h Z;u..._ .
"
j o Prior to the trip the steam rupture
!
produced no changes in reactor power,
;
steam generator pressure, or other
l
primary parameters.
.
I
i
;
' o Normal Post-Trip response observec. 4
.
o Operatinc crew made a cecision to
emergency borate the reactor.
o No significant challenge to the operating
crew.
e
8
.
.
. _ .
.
_
 
_ _ _
_ _ _ _ _ _ _ _
'e
j. Sec6ndary Plant Impact
.
!O
1
_
~w..;gappymast w ? .
-
,
i
l o Damage Assessment- Crew
l
Investigation
:
:
l
!
!
-
I
One MCC was c e-energizec
.
!'O
,
i
Some piaing in the immediate area o"the
!
rupture was bent / twisted.
.
!
l
l
!
!
Wetting of Equipment.
!
i
}
t
!
!
Steam Volume contained in the Turbine
j
Building.
!O
i
!
9
.
 
___ __
___ ______
!
.
l
Sec6ndary Plant Impact
-
,
.
!
4
e
'
'd
.' - .: '
r
: .' ; 4 + , ,. .
o Fire suppression actuated in the
.
:
Turbine Building basement anc
mezzanine.
.
Intermittent ground alarms on a 480V bus
and DC bus #1.
O
No adverse plant effects were observed
due to t1ese intermittent grounds.
0
10
-.
.
. .
.
 
-_
l
Qperational Safety
.
tO
j
Significance
-
.
l
- . 22:= r .x
! o Event effectively mitigatec by the
j
o aerating crew.
[
Minimal effect on the arimary plant.
:
-
l
Minimal effect on t7e secondary plant.
!lO
Minimal reduction in Fire Protection System
l
capability.
i
,
l o Not a significant operating challenge to
1
i
the crew.
l
l
l
:
i
l
lO
!
:
11
i
.-
__
_
_
_
_
 
-
-
---_- _ -
_
.
.
.
e
:~~~ru :. . .
l
~
Maintenance Rule
e
Jim Til s / John Johnson /
-
Ken Dowdy
9
12
'
_
 
---
-_
-
_ _ _ -
--
-_--
.
Maintenance Rule
-
.
Discussion
:
!
y__ .
o Scoping of Systems Under the
Maintenance Rule.
Goal Setting / Performance Criteria
,
requirements
-
o Performance Monitoring of the
O Extraction Steam System.
EC program monitoring effectiveness.
o Use of Industry Operating Experience.
Program Development.
Post-Event Corrective Actions.
o Imalementation of Maintenance Rule
Requirements.
O
13
-.
-
_ _
_ _
.
. . _ . . .
.
.-
.
 
-
_
_
_
__
_
_.
Monitoring Under the
.
.
*
Maintenance Rule
ygn;=_
.
o Sco3ing of Extraction Steam.
Included within scoae of program as part of
t1e Main Feedwater System.
- No failures of Extraction Steam had caused a
plant trip at FCS.
O
- Industry review indicated that Extraction Steam
piping could potentially cause a plant trip.
Classified as Nonrisk Signi"icant
- Based on FCS PRA.
- NUMARC guidance.
Reviewed and Aaproved by an FCS Expert
Panel.
O
14
.
. _ _ -
.
 
_-_
_
_
_ -
- .
,
!
*
!-
Monitoring Under the
.
lU Maintenance Rule(cont)
;
- - _,_
y
, : , , _ u-
_
,.-.c
! o Extraction Steam SSCs were monitorec
l
using Plant Level Performance Criteria.
l
l
NUMARC guidance provides that Plant
l
Level Criteria are ap aro ariate.
~
,
i
Plant Level Performance Criteria
lO
established were:
!
!
- No Plant Trip due to MPFF.
!
l
- No Unplanned Capability Loss due to MPFF.
:
!
- No Safety System Actuations due to MPFF.
System Level Performance Criteria:
- System must be available (100%) when required
for power operation.
g Reviewec and A3provec by Ex3ert
Panel.
15
 
_
_.
_ _
_ _ _ _ _ _ _
- -
Monitoring Under the
-
.
Maintenance Rule / cont >1
\\
1 , f ". r " : * i . i -
o Evaluated against Plant and System
Level Performance Criteria.
Review of SSC failures and maintenance
history was performed from 7/1/92 to
6/30/95.
System monitoring ongoing since 7/1/95. O
No failures prior to the rupture were
identified that exceedec the Performance
Criteria.
o System placec in Category (aX2} of the
rule.
o Reviewed and Approvec by Expert
Panel.
O
16
-.
.
.
.
 
. , . . _ . . ~ . .
!-
l-
Monitoring Under the
.
l Maintenance Rule (cont)
;
a
: =u r_ .
o Piping considerec effectively controllec
l
by Erosion Corrosion Program.
!
l
l
Review incicatec wel-developed inspection
j
prog ram.
O
Proactive replacement of piping prior to
reaching minimum wall thickness.
O
17
 
_ _.-.__
_____
i
-
-
l Status of Extraction Steam .
l Piping After April 21,1997
l
.. _ r: m e t _
.
i
i o Actions taken as a result of the
!
Extraction Steam Line Break.
;
l
Piant Level and Main Feedwa":er System
'
Performance Criteria exceeded.
Cause Determination performed based on
RCA and Failure Analyses.
g
- Identified Failure as MPFF.
- Recommended all FAC-susceptible piping be
placed in Category (a)(1).
o Findings and Recommendations
Reviewed and Approved by Expert
Panel.
o Additional Evaluation.
All plant system piping reviewed.
$
18
.
.
-
.
.
.
 
.
.
l Summary of Maintenance
:o
l
Rule Compliance
-__
,
o Industry operating experience was
properly taken into account. ('" 9.3.3}
4
i
o The extraction steam system classifiec
!
as "nonrisk .significant". ('" 9.3.2}
llO
Guidance. allows monitoring using plant
level criteria.
- No automatic reactor scrams.
- No unplanned safety system actuations.
- No unplanned capability loss factor.
System level aerformance criteria was also
established.
- Requires 100% availability during power
operation.
O
19
 
.
Summary of Maintenance.
.
*
Rule Compliance
.
.
. 1 enre.1. .
,,
o The failure history on extraction steam
was reviewec for a four year period
prior to July 1996. {'" 9.3.3}
.
No problems were detected involving
extraction steam (no through wal leaks ag
no alping below minimum wall).
o Based on failure history, extraction
steam did not require saecific goal
setting and was correctly monitored in
accordance with 10CFR50.65(aX2}.
{'"9.3.4)
e
20
__
.
.
 
_
_
_ _
_ _ .
_ _
_ -
.
.
l Summary of Maintenance
.
;O
l
Rule Compliance
l
. : :n;=rw
:
i o Extraction steam was monitored by the
j
EC program. At the time of Maintenance
l
Rule implementation the EC program
l
was judged to be ef ective.
.
!
;
Proactive replacement of piping
iO
mponents.
!
l
'
Effective in meeting performance criteria
and preventing repetitive MPFF.
Highly susce atible extraction steam
components extensively monitored (70%
sites inspected).
Multia e
ssessments indicated t7e
O
3rogram was an Industry leac er.
21
. _ - .
 
--___--____
_
_
!
-
l Summary of Maintenance.
l
Rule Compliance
*
i
.
___
-
.
.
.-,... ...
..
.
: .
,,, . .. .. .
.
l o The April 21,1997 failure caused both
plant and system level performance
'
!
criteria to be exceeded.
!
!
-
!
!
l
Failure was the initial MPFF.
1
#
l
l
No extraction steam MPFF had occurrec
'
prior to Maintenance Rule implementation.
This is the first known large radius swee a
failure in the industry.
O
22
--
-
-
__
 
__ _______
_ _ _
!
-
l Summary of Maintenance.
lU
Rule Compliance
.
_
l o Based on the failure the following
l
actions were taken:
1
!
A cause determination was performed.
FAC susceptible piping was placed in
l
category (a)(1).
iO
i
Goals were established.
l
Industry experience was again used during
!
both the goal setting phase and cause
determination. (C 9.4.1 and 9.4.4)
,
!
!
; o A determination was made on whether
j
other piping systems within the staae of
l
the Rule were being effectively
j o monitored by the EC program.
!
23
,
1
i
-
.
.
 
;- -
--
-
.
_ . _
- - _
_
;
.
.
l
Conclusion
.-
:
e
: ~x t
_
l o While a Significant Event, the Failure Is
Not, in Itself, a Violation.
!
!
!
' o The Rule Worked As Intended.
~
l
l
l o Extensive Corrective Actions Were
#
l
Taken.
:
l
o Industry Experience was utilized per
NUMARC guidance.
o No Violation of the Maintenance Rule
Occurred.
O
24
-
.
_
.
.
 
a
aru.--n2musa..
s
s -m eu
aesu-asma>sms-ma-ea,nsn2.----.
.u.**-,sa,+ona.
.x.*..
so.<=a=a-
**.um-*-~-.
mm+*-~~a-=sman--=-~,-*-
+ , -
a~~.s-+a----a
-
*
i
.
!.
f
=
lO
.
i
'
=_ .n,
:
!
!
!
l
Assessments and
;
i
.
l
Corrective Action
l0
Update
i
i
!
Joe Gasper
:
O
25
-
 
.
-
_
_ - - - _ _ _ _ - _ _ _ -
.
'
l
Assessments
*
Objective
.
- , - _
-
__.........................,.,.s......
..
l
!
:
!
!
!
To ferret out the root cause
and any other program
,
deficiencies or problems.
4
9
26
.
.
.
-
 
. - -
-
-
- - _ _ _ _ - . - _ _ . _ - _ . _ _ -
I
-
i
.
i
Assessments
-
.
l0
:
!
l
!
-
:. wico, m ,,,, .r. c
;
.
:<..,,n,
.m
.-.,;.
$
i o Initial Assessments
i
-
'
i
;
!
i
~
:
i o Additional Assessments
!lO
:
1
o Corrective Ac: ions
!
4
}
!
,
i
:
i
I
!
27
:
.
_.
.
 
. _ - - . _ _ _ _
-
-
-
_ -
.
.
Corrective
-
Assessments
Actions & G
I
Results
.n ,
- ,
-...; w v.w w.wa - . - 2* gym - - -
er, . . . .
l
+ Lessons
W_=.,
5
-
ong Term
l
+ Corrective
lW-
'
28
. . - -
.
...
-
-
. .
.
.. .
- . .
. - -
-
.. ...
 
_ _ - _ _ _ _ _
_ _ _ _
l
.
-
LAdditionalInspections and
lU
Replacements
l
::
,=
-
l o 23 Sites Inspected
o Replaced because of FAC
l
4th stage sweeps.
6th stage 18 "- 45o elbow - Conservatively
,
l
Replaced - Would Have Reached Design
l O
Minimum Wall in 3 0 3erating Cycles.
o Replaced for other reasons
.
l
6th stage 18"- 45o elbow and pipe - Weld
)
fit-up aroblem aroduced locally high
'
turbulence.
Heater Drain - Three Identical 3 " Pipes
[
downstream of orifices - Visual examination
{
of pipes found no localized wear.
!O
In lusions orlaminations may have causec
l
erroneous UT indications.
;
29
I
. .
.
-
-
-
. .
.-
 
Altrrn Ccrporttiin
Technical Report N . 97152-TR-01
'
Revision 0
i
O
R
=36
0lS;
-
kHgAsq;
A
CI q%
At4
!
$3
R'
$
Q
3
*/
199
g
IN3
.
e.
e
,6
N
SS
y3
-
43
IN
.
gC
R
0474
f
IN
1
P(q 04
>92
gwl,{ ,fj!;
''2
%s,
s
-
Ett 6g
ER
6 D
0
g
I4
6-
IN
E(
'''
%$, $1
$.
No
0A74
4
,,,
9
y
**
N
4-
e,.
%
NOTE: % WALL LOSS CALCULATED FROM too. AT INSPECTION DATE INDICATED.
FORT CALHOUN 4th STAGE EXTRACTION
STEAM LINE TO F.W. HEATERS
,,g;;yggL,
Geometry & Inspection / Replacement
Sununary of 4th Stage Extraction
.
 
-_
-
_
__
._ .
_
i*
-
l
Replaced Components
.
; o
l
i
.
.
.
+ :aguansum;;
-
.
.,..,
.a.
...
. p .
. . .
.
. . . ,
i o Conclusion
,!
.
l
Two sweeps required replacement:
.
l
- The ruptured sweep elbow and
.
l
- The 10" sweep due to FAC (below rninimum
l
wall).
!O
l
Two components showing FAC were
,
l
conservatively replaced (above minimum
l
wall}.
l
l
Four com 3onents replaced for other
:
!
reasons.
i
!
!
l
!
l
30
i
 
.
Update to Assessments
-
.:: : w n
..
o Failure Analysis
o FAC Code Verification
-
o FAC Program Implementation Review
Additional Information Concerning
Replaced Com aonents.
O
'
31
_ - _ _ - _ _ _ _ _ _ _
 
_ _ _ . . _ _ _ _ _ _ - _ _ - _
i-
.
:.
Failure Analysis
.
l
Completed
;
-
. - -
!
l FPI and Altran Concluded:
l
l
i
;
j
The Root Cause was Flow Accelerated
,
l
Corrosion.
l
llO
;
!
There is Evidence of High Velocity Water
DropletImpingement.
;
:
i
Complex hydraulic arofile aroduced large
variation of oxide accumulation and
damage characteristics in failed sweep
elbow.
O
32
.
 
.,.__wa__
4
-*u,.J.
4-m.-am-e-.4.--ie4m
e eis - 4
eal m..a._sm,4u2hh=4e
e.enN J
e
e.Msa-J A eM Aa.-a4-4
-
-,e.cet
-
""ada4.,dezpa.h 4, AA 4 p A se .mm 4-
-
-A4,.e,ide.;W.M_.
@
Water Droplet
-
l
Impingement Craters
*
I
c.
eny g g 4 g
j;g.y...
.
.
.,
-
,
I
i
.
s..
e
c.
, w . wi . . ., ;.,
m 3 o r- - - -- ur p.r. ,. . . . .
.
I
1
j
> T.
''
.g
r.
. . y-p ,m m . .
w.a+
a
1
*
!
Y
QY*.Y.Ti T
~
.
P M. ,. a,e. ,9 P:% . , M -$' M ? G i d g
w.P7p..
u
i
,
,. ., . n. ,,, .- ..m e
-
.
.
..
.
h,$$fy;
_
..
s--
sa
s.
w
L
-
f. v.T"n
m% .-s. p;.
y3
. , . -g ; .
s e -. ' -
-,-T
ky
n . ny -
n
,
: ..
- a s r:w~~. .
-
-
-
gg
. . . ? , '.E;,i kk ... : C%: i '''
~'
-
** t**u ''
!.
;;g <&
', NW; ' "" $
.
y
.
4
+
,,g,,..
. .g
,
s.n..?
+5.
.
%
~O Q L
-
^^ ,
~. .
W .,,
W _ y * ,^- ##. - .,~,g .,; !W:ft
j,
_,
.r
. . . . , .
Ey
' [M Utsc "w a p . ,
' t
1 ~ ' "'s
j - ",; * g; .*
W''-
g
f**
.. . .**q
e .j
a
v.
-*
r-
,=
it%
., '.7^4. . *i '
'6 *;L*%.
,T
T., T T. . ''-
-
*
N,
g .. *J
f=j i * '
.
1
+4
[
, _
g.
y 's [-
h. w.*,j .,
~#
'58
*'
-J2
t
..
- :
-
..
-
.
.ma
-
^ ^
,
.
-
py-
n g-
r; 4
.
2.
-
rj
. 1
. ~. s
.
s.
.g
.
ej
-
. _..
g ,- e % , gj- 4 _'' % . %*a ,
#
n d w-
m, , v.m
s_
.
m
,
1
~;E '* V-
=
.
%
,
y ,a f . . i
W..' . t
.
- - - *
^''T*
$ i.
*k
Re
fy
&,.
-
j
T..
y
s
-
y
. , ' . . . - + %< 4 * * -
T+ *..
..
s -
.. .~ ..
, p ..
;g
.
)
*
.
,
.
;-4, I j f
. : -- -
,]
fC,
'' y
f .** d *
..i.
T-
#
*
. - 'y
* "
<*-;
~
--+
-.
.
$ :,:, . =
h. " ..
'' * *
'. t
'
, ; ', +
,
,
~
y
..
^*
# '
* _A
' J '-
f
. .
,-
_,
].t
. . -n
a.m
*
-
:.*-
-
-a .
...
-g.~
- t
(_,,
-
'
i*
U od*J
~
.
7
, '
M
' %[ % g eWT .
.
':- u x .
* - -
.1 x r
---
urt .
-
y
y
.%
O
33
- -
.. -
.
.
 
_
_ _ _ _
.
_ _ _ _ - _ - -
'*
i
.
l.
FAC Program
.
io
,
!
;
:s u&2.?ttw*m :: . ,.
s
o Code Verification
!
!
l
CHECWORKS Modeling
'
l
1
-
l
BRT-Cicero Mode ing
O Program Implementation Review
!
l
Use of Plant 0 3erating Experience
i
i
!
Site Selection Methodology
l
Use of Industry Operating Experience
i
!
:
i O
i
:
i
34
.
.-
.
_
.
..
. ..
.
-
 
Altran C:rporrti:n
Technical Report
. 97152-TR-01
,
O
#
'
g5 N
%,
-
-
0t 2$>,
N
!\\,
IN
khgk
'#
b4IA
N
s
Ng
'
y
1
<
7
IN
-.
o.
r#
199
4
N
{93
I((,,@.
ing
.
,<gf
. ce2
"4
%$,$,?,hb
lt@
'
&?**
''*"44
*$$,3
9
'
,
,
%Qpp
"''
hO
Cl
?.
**
g
>
o
2
9a7
E
1
OMx
4
NOTE: % WALL LOSS CALCULATED FROM t
AT INSPECTION DATE INDICATED.
FORT CALHOUN 4th STAGE EXTRACTION
,
,
STEAM LINE TO F.W. HEATERS
,,ggg g g ,,
j
Geometry & Inspection / Replacement
Summary of 4th Stage Extraction
.
.
-
-
-
-
 
.
..
l CHECWORKS Modeling.
!o
:
i
I
~
'
. < _
.
.. _ .,
!
o Update and Validate CHECWORKS
l
using 95,93 and 97 inspection data.
j
Parallel trains modeled.
l
Model Verified.
-
j
- Modeling errors corrected.
I
- Irispection data matrices entered for
lO
components.
!
! o Model updated with all data available
j
before 1993 outage.
:
l o Line correction factor.
Value reflects accuracy of predicted wear
!
rate.
Acceptable range between 0.5 and 2.5.
$ age
2
4
6
Factor
2.368
0.370
0.695
35
 
.
9
: CHECWORKS Modeling.
.
e
(Cont)
,
yygny.;
o Measured versus Predicted Wear Plot
shows half of data points outside +/-
50% range.
.
O Conclusions:
e
A good corre ation is not established
between measured and aredicted wear.
CHECWORKS should not be used to
determine the wear status of components
that have not been inspected.
e
36
.-
.
.
-
-
 
.
$
!
BRT-CICERO Modeling
.
lO
.
. ; .2i"'Q
l o Code used by EDF.
!
l o Fourth stage modeled.
o Results
i
~
j
Predicted wear at end of Cycle 16 of 0.429
!
inches.
iO
i
Prediction ~is conservative to observed wall
loss of 0.057 to 0.151 inches depending on
,
l
actual initial wall thickness and a failure
thickness of 0.050 inches.
o Differences between Cicero and
CHECWORKS.
Uses component vs line correction factors.
Calculates individual component prediction
O
uncertainty.
37
 
.
Use of Plant Operating
-
e
Experience
mem_
o RCA Contributing Cause: Incomplete
.
utilization of plant operating experience
Additional replacement data identified
.
Now classed as a root cause
g
o Additional corrective action
Research, maintenance and configuration
changes done before program
implementation in 1988
,
O
38
 
_
-
-
_ _
_
_ _ _ _ _ _ _ _
l
Site Selection Criteria
.
io
!
,
!
'
r
.,um m ,x+.n
-
l o Applied NSAC 2 2L Guidance using
!
data available prior to 1993 outage
l
Inspect Highest Wearing
l
Components.
-
,
!
Shortest Relative Remaining Service Life.
l O
One Component from each parallel train.
i
j
Immediately downstream of control valve
l
and orifices.
i
l
One com aonent in each two phase line of
!
3iping.
:
Industry Experience.
'
Plant Experience.
Replacec components and components
within two pipe diameters of replaced
O
components.
Unusual geometries.
39
.
 
-
_ _ _ __ _ ___ _ _ _ _
Use df Industry Operatin(
Experience
..=~r u-
o Research of CHUG data base identifiec
no additional information and no
information relative to sweeps.
.
o Inspections basec on industry operatir$
experience.
70% of Extraction Steam sites inspected
before 1997 forced outage.
48% of the Heater Drain sites insaectec.
38% of tie Circulating Water sites
inspected.
35% of the Condensate sites inspected.
O
40
.
.
.
\\
 
---- _ _ - _ _ _ _ - _
.
9
FnC Implementation
.
.
l
Review Conclusions
i
~ . r., ., m, ye.
l o Industry experience effectively used.
:
!
}
!
o Plant exaerience not effectively usec.
$ CHECWORKS shou d not 3e used to
predict wear for components that have
not been inspected in the 4th stage line.
o Unique flow conditions may exist in t1e
4th stage line.
O
41
-
_
-
._
 
.
; Current Corrective Actions-
t
e
i
Status
4
~
'
,.
-
.._m
-
,_...s...
o Conduct additional inspections to
develop PASS 2 mode s for
:
:
CHECWORKS.
: o Evaluate on-line radiography for small
bore piping.
e
; o Evaluate replacing high wear piping with
wear resistant piping.
o Evaluate additional moisture traps on
extraction steam piping to reduce wear.
o Provice additional training.
e
42
.
_
.
.
_
.
 
__
_ _
_ _ _ _ _ _ _ _ ___ _ ,
li Current Corrective Actions
!O
l
Status
;
.
.
., ,. .l.,(l ? ~
'
'#
o Work with EPRI to share experiences
l
with the industry.
:
l
l
:
i
~
l
l o Work with EPRI to improve modeling for
l O large radius sweeps.
i
l
!
! o Work with EPRI to better understanc
:
j
effects of oxygen concentration on
j
secondary systems.
!
!!O
!
!
43
;
 
. --
. . _ .
- ---
-_ - _
.-
- _._
-.
.
.
-
.
i
*
O
.
nn
-
'<r<s9'
w .u
%.
. m ,s.,
, . . . , - _ . ,
..
-
1
.
1
i
OE Program
.
Assessment
e
l
Dick Andrews
e
44
.
, _ .
.
_
_
 
._
-
-
-._
.-- _ _ _
.
e
OE13rogram Assessment
.
i O
;
l
'
i..i
~
[
.I. . . '
.
'
'
, . . , .
.,
i o Focus on Use of Operating Experience.
.
'
101 Programs Evaluated.
!
To 3ics Reviewed.
I
- Issues warrant additional Management .
!
attention?
:
i
- Other locations known to be better?
.l0
- is industry experience being obtained and
j
utilized?
j
- is all other needed information being obtained?
l
- Are there barriers to using information obtained?
1
j
- Is information receipt adequate and being
4
properly utilized?
- Are personnel contacts and meeting participation
adequate?
- Have industry experts or peers provided input or
review?
- Is electronic information received adequate and
O
being utilized?
- Is printed information received adequate and
being utilized?
45
 
j
.
.
;
OE Program Assessment . .
e;
..... 2 " ="5 2 *L _
.
,
o Significant Results.
Oversight groups more aware of need to include
1
information utilization when performing audits,
surveillances, and assessment activities.
.
Heightened program owner awareness of ownership
and the need to use external as well as internal
g
information.
Discussion of nuclear operating experience at plant -
morning meetings has been implemented.
s Too many people rely solely on the formal OE
program for their external information.
.
Formal as well as informal industry sources of
information have been identified.
O
46
.
 
_
: O
!
!
. J.ffy c . . ; . . . -- _ ; .
2.. f; , F,'
. ..
-
-
-
..
.<.,.m m m m y w .
. .n
..
-
.
) o 13 of 101 Programs Identifiec for
l
Additional Management Attention.
!
!
Protective Coatings
:
~
Setpoint Contro
-
Design Basis
!O
Radiological Consequences
l
Asbestos Management
l
ASME ISI/IST Program
j
Tagging
Chemical Contro
Hazardous Materials
Nondestructive Examination (NDE)
Bills of Materials
Offsite Dose Calculation, Radiological
O
Effluent Monitoring
M&TE
y
-
-.
 
-
-
. _ _ .
_
_
___
_.
O
;
OE Program Assessment
-
:r -
o Programs identified for Technical
Review.
Erosion and Corrosion Programs
Fire Protection Program
PRA Program
Maintenance Rule Program
SG Inspection Program
g!
n ASME ISI/IST Program
AOV Program
MOV Program
Relief Valve Program
Check Valve Program
9 PM Program
Procurement Engineering Program
Fuel Reload Analysis
EEQ Program
SQUG Program
g
Control of Heavy Loads
Containment H Generation
2
48
 
- _
_
. _ _ _
. _ _ _ _ _ _
. .
j
!
l
!
.
l
i
*
; o
,
4
l
h
' - ~i: l m;%&C%MGV , ifll<
;
.. m.a.u
... . . .
_ m , a m.w. . ..
I
:
!
I
.I
!
!
!
!
!
!,
l
Summar
and
.
,
t
l
OPPD Perspective
!o
;
l
)
'
Sudesh Gambhir
'
!
!
i
l
l
1
l
:
l
5
i
i
,! O
4
0
49
<
i
. - . .- _
 
i
-
Summary and
;
.
*
OPPD Perspective
:
y.n;3
.
: o This Is a Serious Situation Which Must
l
Be Prevented and Not Repeated.
o Significant From an Industrial Safety
.
Standpoint.
o Significant From Plant Availability
Standpoint.
Not a nuclear safety significant event.
- Did not present a significant operational
challenge.
e!
50
 
_ _ -
-
_ _ _
_ _ _ _ - _
_
_
..
;-
Maintenance Rule
.;O
;
Compliance
!
.
: ver _
.
=
i
; o Rule requirements were met:
:i
i
l
Piping was aroaerly monitored.
l
l
Appro ariate aerformance criteria were
j
established.
O
Industry wide operating experience was
utilized.
Monitoring was in place.
- With the exception of 4th stage extraction steam
piping, the program was effective.
o Further Assessment Is Planned to
Evaluate Implementation Against
" Excellence."
O
51
w
 
,
- _
_ _ _ - _ _ _ _ _ _ _ _
.
.
l
FAC Program
.
i
e
;
i
j
Jl2 *2L L.:
i
j o FAC program used for monitoring under
j
the Maintenance Rule.
l o Implementation of FAC program was
!
weak.
-
!
l o Industry experience was factored into
l
the inspecti.on program.
e
!
CHUG Database does not lead to
inspecting sweeps.
-
70% of sites in the extraction steam piping
l
were inspected prior to rupture.
o OPPD was deficient in utilizing are-1988
Fort Calhoun exaerience.
o FAC program limitations (Ref. EPRI doc.
I
NSAC 202, rev.1 page iv).
e
Tab 2
-
52
 
_
_ _ _ _.__
_ . _ _ _ _ _ _ _
;-
.
l.
Assessments
.
i0
'
!
!
.
. w e a m w w o ,e ::
-
.
. .:. n .a
...;
.
~,m.~,-
: .,s. . ,.
.
;
!
!
!
!
:
l
OPPD has conducted
~
!
multiple assessments to
0
!
ferret out the root cause
l
and any other program
i
!
deficiencies or problems.
!.
1
i
l
lO
:
l
53
.-.
_.
.
.
.
.
 
I
'
j
.
-
l
Corrective
-
l
Assessments
Actions & O
;
l
Results
.... _: C =
N"'
. . .
-
+ Lessons
Learned
. Updated
RCA
O
M
Short Term
i
Corrective
E+
o rec
lU4E
i C-
O
i
54
_
. . .
.
. . -
.
. -
. .
.
.
.
_ _ _
 
l
Other Considerations
.
;O
!
!
l
.
..
m
,_ _
.
m v
s o a a ,:.ma,.
! o Corrective Actions
l
l
l
Extensive corrective actions including
i
consideration for generic impact.
.
:
i
,
o Historic Issue
lO
l
Problem occurrec because ofinadequate
treatment o" re alacement 3rior to 1989.
!
!
!
Missed op 3ortunity to inspect in 1985.
o Lessons Learnec
Lessons learned lave been shared with the
O
industry.
55
,
1
.
 
_
. _ _ _
_ _ _ _ _ _ _
- - - - - - -
l.
-
-
4
i
4
i
e
'
.
-l
}
-,
. ~ ~ 4' ?.' i'',7MC . . ' '[ N '
*
*
;
-
.
.
. . , . . . . . . . .
. , . . . . . . . ,
i
;
!
1
>
:
,
l
Closing Remarks
~
,
i
i
e
:
i
I
s
a
;
Gary Gates
i
5
+
J
l
9
56
._
 
1
.
*
1
METHODOLOGY OF MONITORING EXTRACTION STEAM UNDER THE
MAINTENANCE RULE
^
White Paper in response to NRC Inspection Report 50-285/97-09
July 9,1997
1.
PURPOSE,
The purpose of this document is to explain the methodology used to determine
the scoping status of Extraction Steam components, the methods used to monitor
these components, and the basis used for determming the monitoring of these
components.
l
2.
REFERENCES
a)
NRC Inspection Report 50-285/97-09
b)
NEI 93-01 Revision 2, Industry Guidelines for Monitoring the
.
Effectiveness of Maintenance at Nuclear Power Plants
c)
NRC Regulatory Guide 1.160, Monitorine the Effectiveness.o_f
Maintenance at Nuclear Power Plants
d)
NRC Inspection Manual - Inspection Procedure 62706, Maintenance Rule
e)
NUREG-1526, Lessons Learned from Early Imolementation of the
*
Maintenance Rule at Nine Nuclear Power Plants
f)
Questions and Answers from the August 1993 NUMARC Maintenance
Workshops
g)
FCS Program Basis Document, Maintenance Rule
i
h)
FCS Maintenance Rule Implementing Instructions (MRII)
i)
FCS System Scoping Manuals (SSM)
'
j)
FCS PRA Summary Notebook
3.
MONITORING OF EXTRACTION STEAM PRIOR TO THE APRIL 21,
1997 EVENT
a)
Details of Extraction Steam Scoping
The Extraction Steam System (ESS) at Fort Calhoun Station (FCS) is
considered to be within the scope of the Maintenance Rule per
10CFR50.65(b)(2)(iii). These SSCs were included within the scope of the
rule since they can cause a plant trip. FCS had not experienced a plant
<
trip due to failure of extraction steam components, but industry experience
indicated that extrac+ ion steam could cause a plant trip at FCS.
Components within the ESS are monitored as part of the Feedwater
.
d
I
whitel.do::
 
--
--
..
'
.
.
Heaters functional groups.' The components within these functional
.
*
groups are not risk significant according to the plant PRA.2 The ESS is
not a Safety Related system, and there are no safety related functions for
extraction steam listed in the plant Design Basis Documents or the USAR.
Accordingly, the functional groups which include extraction steam are
classed as Non-Risk / Operating functional groups.
b)
Monitoring of Extraction Steam
I
i)
The functional groups containing extraction steam were monitored
using plant level criteria in accordance with 10CFR50.65(a)(2).
Guidance provided by NEI states that non-risk significant /
operating SSCs are to be monitored using plant level performance
criteria. This approach is endorsed by the NRC in Reg. Guide
1.160.'
'
ii)
Components within these functional groups have been monito. red
for failure by the plant NPRDS failure reporting process since
1991, when Revision 4 of the NPRDS Reporting Guidance Manual
(RGM) was implemented. The FCS Maintenance Rule Program
(MRP) adopted NPRDS failure reporting methodologies and
expanded NPRDS guidance to cover all components within the
:
PN
scope of the Maintenance Rule. Revision 5 of the NPRDS
;
Reporting Guidance Manual (December 1994) removed
requirements for monitoring many component types that require
;
monitoring under the Maintenance Rule. As a result, MRP
personnel created a Reporting Guidance Manual that supplemented
the NPRDS RGM and included instructions for continued
'
monitoring of through wall leakage of piping. These instructions
were later incorporated into Maintenance Rule Implementing
Instruction (MRII) -3, Maintenance Rule Failure Renortina. With
the cessation of NPRDS reporting, MRII-3 was revised to include
'
all necessary guidance for Maintenance Rule failure reporting, and
the NPRDS RGM is no longer used by the MRP as a reference.
iii)
In accordance with NEI guidance,5 a review of the failure history
;
of functional groups including extraction steam SSCs was
!
performed. This review consisted of two tiers. One was the
' Ref. i) Volume 15, Main Feedwater. Tab 10, LPA(B)HTR IPA (B)HTR HPFWHT - Feedwater
4
;
Heaters
2
,
Refj) 9.133.F.
' Ref. b) 93.2
d
i
Ref. c) 1.73
,
V
' Ref. b) 933
2
whitel. doc
 
'
.
.
review of existing plant specific NPRDS and Maintenance Rule
*
data from CHAMPS. The second tier involved the review of
Maintenance Work Documents from July 1,1992 through June 30,
1995 to determine if additional failure reporting other than that
already contained in CHAMPS was required. It should be noted
that NEI guidance only requires the licensee to perform a review
of history for a maximum of three years prior to the
implementation date of the rule.6 There were no problems
detected involving extraction steam that would require goal
setting.7 Based on these reviews, it was determined that the
functional groups containing extraction steam SSCs did not require
i
goal setting and were being correctly monitored in accordance
with 10CFR50.65(a)(2). While NEI guidance was used to make
this determination, the methodology used by FCS to determine the
proper monitoring level of extraction steam SSCs is dso discussed
in NUREG-1526 and endorsed in Reg. Guide 1.160.30
iv)
Extraction steam piping, as well as piping from other systems,"is
monitored by the plant's Erosion / Corrosion Program (ECP). The
ECP sets individual, component level performance criteria for
piping covered within its scope. The FCS MRP does not set
individual, component level performance criteria, but makes use of
existing programs as allowed by law" and NEI guidance. 2
^
Effectiveness of the ECP is monitored by the MRP using two
methods. These methods are discussed below:
a)
Failure Reporting Process - Through wall leakage of piping
is considered a component failure' per MRII-3. As such, a
failure investigation must be performed and a report
i
generated into the CHAMPS database when a leak occurs.
Since extraction steam SSCs are monitored at the plant
level, a non-catastrophic failure (leak) of an extraction
l
steam line would not exceed MRP performance criteria
unless plant level performance criteria (as described in
MRII-2, Settine Performance Criteria) were exceeded.
However, MRII-5, Component I ailure Analysis, allows for
the elevation of an SSC to monitoring under
.m
' Ref. b) 733
' Ref. i) Volume 15, Main Feedwater. Tab 15. Initial Performance Assessment
* Ref. b) 9.2.4
' Ref. e) 2.4.1
'' Ref. c) 1.9
" Ref. c) B Use of Existing Licensee Programs
" Ref. b) 7.0
v
" Ref f) Appendix C, Section 12, Question 50
3
whitel. doc
 
- _ . _ _
.
_
_
_
._- .
. _ _ _
_ _
.
_
'
.
.
10CFR50.65(a)(1) even if performance criteria is not
.
^
exceeded, if the situation warrants.
b)
A catastrophic failure of extraction steam piping would
cause plant level perfonnance criteria to be exceeded (if
,
deemed an MPFF), or would exceed system level criteria
(Availability of 100% while power operation is desired)
even if the failure was not mairtenance preventable." In
effect, MRP performance criteria adequately monitors the
ECP since piping is monitored not only at the plant level
(as allowed by law), but also at the system and, to some
extent, the component level.
v)
A detailed analysis of the ECP by the MRP was not required as
part of the maintenance rule implementation effort. NEI 93-01
states:
*
Utilities can utilize their existingprogram results to
support the demonstration that SSCperformance is being
effectively controlled through preventive maintenance. Jf
verformance monitorine indicates that SSC verformance is
unacceptable. then the cause determination (Section 9.4.4)
e
verformed when SSC verformance is unacceptable should
correct any eauipment or vrogram deficiency. "
The FCS MRP was monitoring the effectiveness of the ECP, and
when the ECP was found to be ineffective in ensuring the
performance of extraction steam piping, a cause determination was
performed and goals were set as required by the rule.
c)
Conclusions
The following conclusions can be drawn about the status of the ESS prior
to the steam line break of April 21,1997:
i)
Extraction steam SSCs were correctly within the scope of the FCS
MRP.
ii)
Functional groups containing extraction steam SSCs were
correctly included for monitoring under 10CFR50.65(a)(2) based
on a review of maintenance history.
" Ref. i) Volume 15, Main Feedwater. Tab 2, Main Feedwater System
y
" Ref. b) 7.0
4
A v :. doc
 
_
_
'
.
.
iii)
A failure such as that occurring on April 21,1997 would have
-
^
caused both plant level and system level performance criteria to be
"
exceeded. MRIl-6, Placement of SSCs into Category (a)(1) or
(a)(2), would have prompted a cause determination as required by
law and NEI guidance. Such a cause determination would cause a
review of the effectiveness of existing programs as well as current
, performance criteria."If the ECP was fcand to be ineffective,
functional groups containing extraction steam SSCs would be
monitored under 10CFR50.65(a)(1) until effective corrective
l
action was taken. This meets the requirements of Reg. Guide
1.160" and NEI 93-01."
4.
MONITORING OF EXTRACTION STEAM AFTER THE APRIL 21,1997
EVENT
l
a)
Actions Taken Directly as a Result of the April 21,1997 Event
,
i)
The failure of a large radius sweep in the fourth stage extractio'n
line on April 21,1997 caused operations personnel to manually
trip the reactor. The failure of the pipe is classed as an equipment
failure per MRP failure reporting guidance." This requires an. '
investigation of the incident and the generatior, of a failure report.
'S
Failure reporting procedures require that a determination be made
.
as to whether the failure was maintenance preventable or not.
Industry experience is also to be used when making this
^
determination.20 The MRP reviewed the preliminary root cause
analysis (RCA) performed by plant personnel. The preliminary
root cause analysis determined that there were several
programmatic deficiencies in the ECP, and that the failure could
have been prevented. The failed section of piping was also sent
,
'
out for a failure analysis to two different vendors. Completion of
the MRP failure report is pending the results of these failure
analyses.
The failure of the piping caused a loss of the functions.1 groups
'
containing the failed pipe. The failure also caused a plant trip.
MRP personnel had intended to wait until the completion of the
failure analyses before performing a cause determination. This
would be necessary in order to determine if the failure was indeed
" Ref. h) MRII-6. Placement of SSCs into Cateeory (aYl) or (aV2),5.6
" Ref. c) 1.9
" Ref. b) 9.43 and 9.4.4
" Ref. b) MRIl-3djaintenance Rule Failure Reportine. 6.2
v
** Ref. h) MRII-3. Maintenance Rule Failure Reportine.14.29
5
whitel. doc
 
..
.
. -
..
*
.
, .
an MPFF as stated in the preliminary RCA and therefore a
_
violation of plant level criteria. However, since system level
availability performance criteria were also exceeded (which are
applicable even if the failure is not an MPFF), a cause
2
determination was performed as required by NEI guidance and
Reg Guide 1.160.2 This cause determination concluded that the
-
ECP, in its present state, did not adequately monitor Flow
Accelerated Corrosion (FAC) suseptible piping per
10CFR50.65(a)(2). Therefore goal setting would be required
under 10CFR50.65(a)(1). Corrective actions already completed
were reviewed, and goals were set to insure completion of
additional corrective actions. The question of monitoring of the
ECP was asked of the NRC at an industry conference and the
following answer was given:
,
Programs are not in scope; only SSCs andperformance. . .
A firstfailure ofa SSC within the scope ofthe Maintenance
.
Rule due to errosion / corrosion would require an effective
-
cause determination and corrective action. A .second
l
failure ofthe same kind uvuld require goal setting and
monitoring.
;
.
^
In accordance with the above, the ECP was not placed into
category (a)(1); rather all FAC suseptible piping is currently being
monitored in accordance with 10CFR50.65(a)(1).24
25
i
ii)
In accordance with applicable guidance , the plant made use of
industry experience during the goal setting phase of the cause
'
l
determination. Personnel from other utilities were part of tha ECP
self assessment, and other industry experts were called upon to
assist with the failure analysis and to provide advice for program
'
improvements.
b)
Other Actions Taken as a Result of the April 21,1997 Event
.
A review was performed by MRP personnel to detennine if other piping
systems currently within the scope of the rule were being effectively
monitored by the ECP or by other preventive maintenance (PM)
2
*' Ref. b) 93.4
22 Ref. c) 1.7.1 Cause Determination
** Ref. f) Appendix C, Section 16, Question 6
'' FCS MRP Cause Determination alfile\\cc249705 Revision 0
28
v
Ref. b) 9.4.1
6
whitel. doc
 
_
_
__
_
_
. . _ _ _
.__
\\
*
_
]
\\
-
programs.26 This evaluation resulted in the piping for three additional
-
"
^'
,
systems to be transferred from rnonitoring under 10CFR50.65(a)(2) to
!
I
monitoring under 10CFR50.65(a)(1). While these systems were being
monitored under system and functional group level performance criteria,
no evidence could be found that piping was being specifically monitored
.
'
in any way other than the MRP failure reporting process. It should be
noted. that monitoring of these systems was not ineffective since the
failure reporting process had already detected piping problems that had
"
contributed to unavailability that lead to the Instrument Air Compressors
being placed into category (a)(1) in January 1997.27
4
5.
FINAL CONCLUSION
.
Upon reviewing the documentation listed in Section 2 of this paper, and
j
reviewing the monitoring of extraction steam piping prior to the April 21,1997
event as well as corrective actions taken after the event, MRP personnel must
conclude that this event did not constitute a violation of 10CFR50.65. This .
conclusion is based on the following points:
;
a)
The SSCs involved were classed as non-risk / operating SSCs and
)
therefore monitoring at the plant level was appropriate.
A
b)
The Maintenance Rule only required a three year review of maintenance
history to determine the level of monitoring necessary for these SSCs. No
problems were detected during the review period. In addition, the
proactive replacement of several extraction piping components by the
,
ECP led MRP personnel to believe that the ECP was successful in
detecting piping deterioration prior to failure. As such, there were no
grounds for goal setting for these SSCs prior to the April 21,1997 event.
This is in accordance with 10CFR50.65(a)(2) which states " monitoring as
specified in paragraph (a)(1) ofthis section is not required where it has
been demonstrated that the performance or condition ofa structure,
system, or component is being efectively controlled through the
performance ofappropriate preventive maintenance, such that the
structure, system, or component remains capable ofperforming its
intendedfunction. "
c)
The purpose of the Maintenance Rule is to monitor the effectiveness of
preventive maintenance and take appropriate corrective action when
preventive maintenance is found to be ineffective. Paragraph (a)(1) of the
rule requires that "each holder ofan operating license . . . shall monitor
'' FCS MRP Analysis efar\\cc249706
'' FCS MRP Cause Determination alfile\\l4069701
v
7
whitel. doc
 
i
*
l
?
e
I
i
'
the performance or condition ofstructures, systems, or components,
~
against licensee-established goals, in a manner sufficient to provide
reasonable assurance that such structures, systems, and components .
are capable offulfilling their intendedfimetions. Such goals shall be
established commensurate with safety and, where practical, take into
account industry-wide operating experience. When the performance or
conclition ofa structure, system, or component does not meet established
goals, appropriate corrective action shall be taken. "
d)
Extraction steam piping at FCS was being monitored by the MRP at a
level that was in accordance with NEI 93-01, Revision 2, as endorsed by
NRC Reg. Guide 1.160. When these performance criteria were exceeded,
comprehensive corrective actions were taken and all requirements of NEI
93-01 were performed as to the re-catigorization of the SSCs involved.
e)
Industry operating experience was taken into account during the initial
scoping of extraction steam SSCs, and this experience resulted in the ,
inclusion of extraction stream within the scope of the Maintenance Rule.
Industry experience was also taken into account in the setting of goals
after plant level performance criteria had been exceeded.
f)
The failure in question is not a repeat MPFF. The failed component is a
large radius sweep, and there is no available industry information
m
indicating that these components are susceptible to accelerated FAC,
Preliminary failure analysis results also indicate that the flow rate within
the affected piping was considerably higher than expected due to the
configuration of the piping. NEI 93-01 defines MPFF and repeat MPFF as
follows:
An initial MPFF is thefirst occurrencefor a particular SSCf r
which thefailure results in a loss offunction that is attributa!!e to
a maintenance related cause. An ininial MPFFis afailure th:
would have been avoided by a maintenance activity that has n, t
been otherwise evaluated as an acceptable result (i.e., allowed sa
run tofailure due to an acceptable risk).
A " repetitive" MPFF is the subsequent loss offunction (as defined
above) that is attributable to the same maintenance related c.=?
that haspreviously occurred (e.g., an MOVfails to cine because a
spring pack was installed improperly -- the next tir.e this MOV
fails to close because the springpack is installed 'mproperly: the
MPFF is repetitive and the previous corrective c : tion did not
preclude recurrence). A second or subsequent loss offunction that
resultsfrom a different maintenance related cause is not
v
considered a repetitive MPFF (e.g., an MOVinitiallyfails to close
l
8
whitel. doc
 
i
.
.
1
.
because a springpack was installed improperly - the next time it
^
fails to close, itsfailure to close is because a set screw was
.
improperly installed: the MPFF is not repetitive). "
.
The current failure is the first functional failure of an extraction steam
pipe at FCS due to an ineffective ECP. It is also the first failure of a large
radius sweep due to FAC during the historical monitoring period required
;
by the rule. As such, it is an initial MPFF and not a repeat MPFF. Gther
i
extraction steam components (small radius elbows, reducers, etc.) were
being comprehensively monitored by the plant ECP, in part due to the
available industry operating experience considering these components to
be a problem. Since this is not a repetitive event, the failure in question is
not a violation of 10CFR50.65.
.
#f'%
-
.
L
g
'' Ref.b) Appendix B
9
whitel. doc
.
 
*
-
,
T;;tra Engineering Group, Inc.
t toucomeacsw sweet suite 800
r sephone (seos est 4622
e
Westogue, Conneco' cut 06069 USA
Fa phone (860) SS15524
,
r.
1
l
July 17,1997
97-FCA-203
Mr. Joseph K. Gasper
Omaha Public Power District
5 miles North of Fort Calhoun NE on Highway 75
Fort Calhoun Station
P.O. 399
,
Fort Calhoun, NE 68023 USA
Dear Joe:
Encimed please find the final version of Tetra Engineering Report TR-97-009 entitled " Fort
Calhoun Flow Accelerated Corrosion Assessment of Extraction Steam Line".
.
Sincerely,
Frederick C. Anderson
''A
Vice President Engineering Services
e-mail: FAnderson_ Tetra @compuserve.com
-
'
v'
Engineering & Services forIndustry
 
. -
-
--
.-
. -
- - -
_ _ .
_ _ - - -
.
*
_
l*
-
-
I
-
TR-97-009
Fort Calhoun
Flow Accelerated Corrosion
Assessment of Extraction Steam Line
.
cT('g
Prepared For Omaha Public Power District
By Tetra Engineering Group, Inc.
July 17,1997
.
W Tetra Engineering Group, Inc.
USA: 110 Hopmeadow Straet, Suite 800, Westogue, CT, 06089 (1).860.651.4622
France:Immeuble Petra B, it.P. 272, 06905 SOPHIA AN11POUS (33).4.92.96.92.54
-
 
.
. _.
.
. . .
.
.
.
. . - - .
_
-.
-
. .
- -
W revo engineenne croup. Inc.
'
:
,7%
Contents
.
e
'
.
Introduction
1
,
BRT-Cicero Code
2
'
Fort Calhoun 4* Stage Extraction Steam Line
4
Comparison of BRT-CICERO to CHECWORKS
5
7
Conclusion
-
References
8
l
!
i
,
!
Appendix A
9
: a,
--
I
l
<
f
1
l
l
.-
,
i
TR-97 009 Fort Calhoun FAc Assessment of Extraction Steam Line
Contents e i
.
 
L r,,,, ,,,,
, a,c
, c.
-
>
-
-
Introduction
On April 21,1997 a sweep elbow in the 4* stage extraction line of the Fort
Calhoun Nuclear Power Plant ruptured. The cause of the rupture was determined
to be flow accelerated corrosion.
A review of the flow accelerated corrosion program at Fort Calhoun was initiated
following the rupture. One facet of this review is a attempt to determine why the
existing program failed to identify the thinned component piior to rupture. The
Fort Calhoun program used the EPRI CHECWORKS computer code to identify
potential thinned components for inspection. An alternate methodology was
developed by Electricit6 de France for predicting flow accelerated corrosion in
power plant components. The EdF code is entitled BRT-CICERO.
This report contains the results of a BRT-CICERO analysis of the Fort Calhotm
4* stage extraction steam line.
een
-
l
.
l
TR-97 009 Fort Calhoun FAC Assessment of Extraction Steam Line
introduction .1
_
 
_
.
___
. _
_ _
'
Tstra Engine 1nng Group, Inc
4
-
-
.
BRT-Cicero Code
The BRT-CICERO' code was developed by M. Bouchacoun and F. N. Remy of
EdF/SEPTEN and J. de Toni of EdF/CNEPE. The intent of the development of
the code was to provide a centralized method for predicting and controlling flow
accelerated corrosion at the 54 nu: lear plants operated by EdF. The main
objectives of the code are the removal of the likelihood of a pipe rupture and the
reduction of maintenance program costs. The code is currently used by EdF to
l
diagnose the state of piping in the secondary system of a plant, assess component
lifetimes, prepare inspection campaigns, optimize replacement and repair
;
l
strategies, and document analysis,
The code provides a database function for the codes and standards used in the
'
plant construction, an inventory of the lines and elements, plant operating history,
plant chemistry history, and inspection results. Algorithms used to predict flow
accelerated corrosion of components are based principally on a modified form of
the Sanchez Caldera model a supported by testing" performed in the CIROCO
l
2
test loop run by EdF and feedback from operating plants. The basic predictive
~'
algorithm used in the BRT-CICERO code to determine the rate of FAC is as
follows:
i
'
FAC = f(Cr) * f(6) * (C,* - C.,)
(1)
,
g,
0.5 * ( k + D )i
<
Where:
f(Cr)
Alloy Composition Factor
f(0)
Oxide Porosity Factor
C,y
Equilibrium Soluble Ferrous Ion Concentration
C.,
Bulk Soluble Ferrous Ion Concentration
k
Mass Transfer Coefficient
6
-
Oxide Layer Diffusion Factor
D
,
The Alloy Composition Factor is a function of the Chromium, Molybdenum, and
Copper concentrations in the material ofinterest. BRT-CICERO assumes an
average alloy composition if no specific information is available. The average
values are based on extensive testing by EdF of a large number of heats of carbon
steel material. EdF also tests alloy composition of each component inspected as a
. . .
TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam Line
BRT-Cicero Code . 2
 
'
Tetra Engineenng Group, Inc.
i
~
matter of policy. This removes a considerable uncertainty in the FAC rate
<
1
.
predictions which would otherwise be present if the alloy composition is
unknown.
The Oxide Composition Factor and the Ferrous Ion Concentrations are functions
of pH and temperature. The Mass Transfer Coefficient 's a function of velocity
and the oxide Layer Diffusion Factor is a function of temperature.
Equation 1 applies to straight pipes. For elbows, tees, aad other geometric
discontinuities a geometry factor is applied. This geometry factor is a function of
,
the Sherwood number and accounts for the increased mass transfer as a result of
the discontinuity.
5
The BRT-CICERO code is used by first constructing a plant database. The plant
database consists of all susceptible lines modeled from isometric drawings plus
j
data such as construction code, design conditions, and operating conditions.
{
Additional information which must be entered includes pipe material properties,
nominal wall thickness plus tolerances, plant operating history, water chemistry
,
history, etc. Structural margins are then determined in order to quantify the
available wall for acceptable FAC degradation. Wear calculations are tlien
performed on all components and predicted wall thickness with associated
j
uncertainties determined for each component.
l
,
The code is used to assist in the selection of camponents for inspection.
l
Components can be classified in terms of the margin between projected thickness
and design thickness, the wear rate, or the time to minimum required wall
thickness. The code can then be used to determine the minimum inspection
frequency for the component.
When a component is inspected the 'UT information is entered into the code md
measured wear determined. Projections of the future wall thickness are based on
the observed wall thickness.
,
!
i
l
i
i
i
l
TR 97-009 Fort Calhoun FAC Assessment of Extraction Steam Line
BRT-Cicero Code . 3
!
;
 
__
_
. _
-__.. - .. _-__ -
.--
-
-
. -
. -
'
Tatra Enginernng Group. Inc.
-
.
.
Fort Calhoun 4th Stage Extraction Steam
Line
'
The BRT-CICERO code was used to assess the rate of FAC in the failed section
of the Fort Calhoun 4* stage extraction steam line using the data provide in
appendix A. Only the section of the line from the nozzle to the first tee was
considered. The following assumptions were made:
1. The Unit was assumed to be at 100% power for the entire 145000 operating
hours.
2. Chemistry was assumed constant at a pH of 9.44,
3. Fluid conditions in the line were:
'
Enthalpy 2.64x10' kjoule/kg
Flow
36.9 kg/see
Pressure
18.961 bar
4. Chromium content of the failed sweep elbow was 0.068%.
.
5. The radius of the sweep elbow was 1.5 m.
Based on the above input assumptions, the BRT-CICERO code projects a wall
loss of 10.9 mm or 0.429 inches for the 145000 operating period'. This compares
with the range of possible initial wall thickness of the component of 0.328 to
0.422 inches. The range ofinitial wall thickness reflects the nominal wall
thickness of the component *12.5% for the as procured tolerance.
The wear projected by the BRT-CICERO code is conservative to the observed
wall loss by anywhere from 0.057 to 0.151 inches depending on the actual initial
thickness of the component and assuming the rupture occurred with 0.050 inches
of wall remaining. This is reasonably good agreement given the assumptions
regarding operating conditions and chemistry.
1
)
TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineFort Calhoun 4th Stage Extraction Steam Line = 4
I
 
..
.
.. - -
-
-
_ _ _
..
-.
.- - . -
-.-
-
'
& Te*rs Engineenng Group. Inc.
'
.
.
!
Comparison of BRT-CICERO to
CHECWORKS
-
'
l
The CHECWORKS model of the Fort Calhoun 4* stage extraction line was not
;
available for review, therefore a detailed comparison of the CHECWORKS and
BRT-CICERO models could not be performed. However, some observations can
be made with regard to the general calculational approaches. For the purposes of
this report, only observations pertinent to the determination of the 4* stage
extraction line failure are provided.
There are three main differences in the approaches used by the two codes which
may have an effect on the prediction of wear in the 4* stage extraction line. These
differences are related to the use of a line correction factor, the treatment of alloy
content, and prediction uncertainties.
CHECWORKS employs an adjustment factor termed the "Line Correction
j
Factor"in the " pass 2" wear calculation. This correction factor is intended to
adjust the predicted wear rate of all components in a line by considering the
'
differences between the predicted wear and the measured wear for the components
,
on that line that were inspected. This has the effect of adjusting the predicted
'
wear of one component based on the measured wear of a different component,
essentially broadening the use oflimited inspection data. This is useful provided
that all components in the line are behaving in a similar fashion and the inspected
components contributing to the line correction factor are carefully selected.
The BRT-CICERO code does not employ a line correction factor. Inspection data
from one component is not :xtrapolated to other components. Inspection data is
used to adjust the predicted wear for that particular component only.
Both codes use a default value for alloy content of a component when no
information is available. CHECWORKS assumes a alloy content value of 0.0%
for carbon steel components, while BRT-CICERO uses an average value. The
0.0% value assumed by CHECWORKS would be conservative when calculating
the wear rate of an individual component. However, when coupled with the use
of a line correction factor, a potentially non-conservative scenario may occur. A
non-conservative prediction of wear could occur when a limited number of
examinations are performed on a line and the components selected happen to have
unmeasured alloy contents greater than the component ofinterest. The inspected
components would have lower wear rates due to the alloy effect. When the line
correction factor is used and wear for the line is determined, the projected wear
rate for the low alloy content component may be non-conservative.
.a
TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineComparison of BRT-CICERO to CHECWORKS . 5
-
 
bN ras engsneenne aroup. Inc.
'
.
Uncertainty in the prediction of wear is explicitly addressed on an individual
component basis in the BRT-CICERO code but not in the CHECWORKS code.
m
THE BRT-CICERO output provides an average wear value plus upper and lower .
bound values. The uncertainty in the predicted wall thickness is based on the
uncertainty in the initial wall thickness, which is typically 112.5%, plus
uncertainty in the alloy composition, uncertainty in UT measurements, and
uncertainty in the wear calculation. If a baseline inspection is performed or once
the comp' nent is inspected during service, the uncertainty in the initial thickness
o
is eliminated. Thickness projections and associated uncertainties are "re-zeroed"
j
from the inspected wall thickness measurement. This can not be done easily in
CHECWORKS for components that have no initial baseline and that are inspected
after an initial operating period. Instead, a nominal wall thickness is assumed and
wear is emnulative c,ver the life of the component.
Similarly, the contribution of the alloy uncertainty can be eliminated in the
BRT-CICERO code by the performance of an in-situ alloy analysis.
CHECWORKS also has the capability to record and apply the allow composition
j
should the composition of a component be determined. It does not, however,
'
explicitly address the uncertainty of unknown alloy compositions on a component
basis.
!%
.
-
..-
TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineComparison of BRT-CICERO to CHECWORKS e 6
 
. .
. _.~..
. . . . _ .
_ ._ - . _ . _ _ _
_ _ . _ . _ _
. _
..
. _ _ _ _
._ .
N ntre ene,neenne amuo. anc.
'
l
.
-
.
:
Conclusion
.
4
The BRT: CICERO code was used to model the failed section of the Fort Calhoun
.
!
4* stage extraction steam line. The predicted wall thinning from the code was
8
conservative to the observed degradation but still within reasonable agreement.
More important is that the predicted time to failure for this component would have
>
l
been less than the observed time to failure.
I
It should be noted that the exact measured chromium content was used in the
determination of the wear for this component. This in ormation was not available
r
;
prior to the rupture. The value of 0.068% is somewhat less than an average of
m
approximately 0.16W and may have contributed to the high wear rate. If an
l
average value for chromium was used in the BRT-CICERO code a somewhat less
conservative wall loss would be predicted. This analysis was not done, but it is
:
likely that the code would have projected a failure of this component in time to
1
l
avoid the actual rupture.
l
;
i
:,
}
-
a
k
;
:
s
J
TR-97-009 Foot Calhoun FAC Assessment of Extraction Steam Line
Conclusion * 7
 
. . . . _ _ .
.
.
_.
.-.
_ . _ . _ _
.
. _ .
.
. . _
-
W retis engineenne roup, Inc.
'
c
.
; .
.
.
,
e
:
References
I
]'
l'.
Bouchacourt, M.,"BRT-CICERO 'A Software for Controlling Flow
Accelerated Corrosion' - User Manual," Revision A, EdF, Lyon, France, June
i
:
1995.
2. Sanchez Caldera, L. E., "The Mechanism of Corrosion-Erosion in Steam
j.
Extraction Lines of Power Stations," Ph.D. Thesis, Massachusetts Institute of
Technology,1984.
3. Cragnolino. G., Czajkowski, C., Shack, W. J., " Review of Erosion-Corrosion
j
j
in Single Phase Flows," NUREG/CR-5156, April 1988
i
4. Ducreux, J., "The Influence of Flow Velocity on the Corrosion-Erosion of
,
Carbon Steel in Pressurized Water," Water Chemistry 3, BNES, Lond'on,
l
1983.
5. Berge, P, Khan, F.," Corrosion-Erosion Des Aciers Dans L' Eau et la Vapeur
;
'
Humide," R6 sums et conclusion de la reunion de sp6cialistes, Mai 1982.
i
6. M. Bouchacourt E-Mail to F. Anderson," Transmittal of BRT-CICERO code
,rs
results", June 2,1997.
]
I
i
7. Jonas, O., " Erosion-Corrosion of PWR Feedwater Piping Survey of
:
Experience, Design, Water Chemistry, and Materials," NUREG/CR-5149,
;
March 1988.
4
1
:
1
;
,J
TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam Line
References . 8
 
..
.
. _ _ _
. _ .
._
-
.
.
i
s
.
,
i
e**
OEP7blicP7,verD$
~
"
444 South 16th Street Mall
Omaha NE 68102-2247
June 4. 1997
LIC-97-0087
i
U.S. Nuclear Regulatory Commission
Attn: Document Control Desk
'
Mail Station P1-137
Washington. D.C.
20555-0001
References: 1.
Docket No. 50-285
2.
LER-97-003 Manual Reactor Trip Due to a Steam Line. Rupture
SUBJECT:
Assessments Related to the Extraction Steam Line Rupture of April
21, 1997
l
'
As committed in the May 5. 1997 Public Meeting, please find attached the
assessments completed in response to the extraction steam line rupture of
)
April 21. 1997.
These documents contain Omaha Public Power District's (OPPD)
internal findings and recommendations concerning the Extraction Steam Line
,
Rupture event that occurred at Fort Calhoun Station (FCS) on April 21, 1997.
OPPD's corrective actions for this event are listed in LER-97-003.
For the
purpose of providing additional detail to the NRC the corrective actions in
LER-97-003 are expanded upon in Attachment 1 of this correspondence.
However.
'
these specific actions may change as OPPD continues to review and improve its
program and are not meant as additional commitments.
At this time, the failure analysis of the ruptured elbow has not been
received. This item will be sent at a later date.
Please contact me if you have any questions.
Sincerely,
y&
S. K. Gambhir
Division Manager
Engineering and Operations Support
v
c5.5124
Employment with Equal Opportunity
 
-
_ _ _
_._
_ _ _ _ _
- _ . _ . _ _ . .
_
_
_. .
._
, .
_
!
l.
s
Attachments 1.
Additional Information on Commitments to the NRC for
Co~rrective Actions Listed in LER-97-003
,
2.
Fort Calhoun Station Root Cause and Generic Implications
'
Report Fourth Stage Steam Extraction Line Rupture CR
199700445 Revision 0
,
3.
Damage Assessment Report for the Break in the Extraction
j
Steam Line Revision 0. May 3, 1997
,
4.
Fort Calhoun Station Erosion / Corrosion Program Assessment
'
Report, dated May 2. 1997
i
5.
Fort Calhoun Station Self Assessment Erosion / Corrosion
Program Team Findings, dated May 6. 1997
SKG/ddd
.
j
c:
Winston & Strawn
E. W. Merschoff, NRC Regional Administrator. Region IV
'
'
'
J. L. Shackelford. Senior Reactor Analyst. DRS
l
L. R. Wharton. NRC Project Manager
.
W. C. Walker. NRC Senior Resident Inspector
1
.
.
"
g
$
"
>
1
a
)
,
.
~.)
, _ - .
 
_ ._ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ ._. _ ._ _ . _ _ __ _ ..- _ __. _ . ____ - _._ -__ ..__ ____
j
-
.
!
;
i
Attachment 1
'
-
i
;
-
l
Additonal information on Corrective Actions
!
Listed in LER-97-003
:
$
i
4
-
d
'!
t
i
1
'
r
$
l
4
1*
.
4
.
J
r
i
.I
}
i
i,
4
,
!
"
j
..
.
!
!
1
i
-
.
:
i
;
.
.
!
\\
l
1
~
i
i
1
!
!
i
:
.
-
/
4
4
4
.
6
nv-
e,
,
v
-
- - - - - - -
- - - - ,
..-
-,e-
,
, -.
-e,,-
 
. _ . . _ . _ _ _ _ _ _ . _ _ - _ . _ _ _ _ _ _ . _ . _ . . _ _ _ . _ _ _ . _
j
-
-
.
!
d
l
Attachment 1
'
-
-
b
Additional Information on Corrective Actions Listed in LER-97-003
)
!
In LER-97-03 OPPD committed to " Revise the Erosion / Corrosion Program Plan.
;
controlling procedures and modules to be consistent with industry standards.
!
This revision will include upgrade of the implementing procedures to be
consistent with industry standards (e.g. NSAC 202L. Rev.1), development of
;
susceptibility documentation and requirements for use of current industry
i
experience. This will be completed by the beginning of the 1998 Refueling
i
j
Outage." This commitment will address the following issues identified in the
;
self assessment:
l
1.
Procedures should include more specific guidance on how Outage
i
inspection locations ~are chosen.
i
1
i
2.
Measured wear determination process should incorporate the following
;
industry practices:
.
1
-
'
j
a.
Use of accepted practices to determine lifetime component wear
j
(circumferential band, moving blanket, point to point).
i
l
b.
Clarify the use of engineering judgment relative to wear
determinations.
j
-
,
.
;
c.
Process for incorporating measured wear into CHECWORKS models.
-
l
.
.
,
;
3.
Sample expansion process should be revised to align it with industry
i
standards.
Specific changes include:
l
-
i
a.
Clarify wording for small bore piping.
l
b.
Add requirement for inspecting upstream of expanders / expanding
elbows.
2
c.
Clarify that pressure / temperature exemption only applies to raw
water systems.
d.
Clarify that expansion is to parallel components in each train.
~
e.
Define the terms " component" and " highest wearing".
7
Specify highest wearing components in the same train.
4.
A document is needed to describe and control the identification of
susceptible systems.
.
1
 
-
-
-- .
._
.
_.
-
. - - - -
..
.
_
_ -
-.
*
.
;
.
5.
A document is needed to describe and control the evaluation of
~
'
- . .
susceptible systems.
:
,
6.
Documentation should be provided when grid refinement.or scanning is
3
performed when significant erosion / corrosion is found.
i
7.
The inspection data evaluation for components with PASS 2 CHECWORKS
.
analysis should consider current predicted wear rates.
'
l1
8.
The program should employ verification of many elements and a formal
process for changing program elements needs to be established.
(Examples: CHECWORKS model changes. Erosion / Corrosion Program General
Information Table. Erosion / Corrosion Program Technical Data Review.)
i
.
Also in LER-97-03 OPPD committed to " Revise and verify the Fort Calhoun
CHECWORKS models consistent with industry standards by December 31. 1997."
i
This commitment will address the following issues identified in the self
assessment:
4
.
.
1.
Current plant CHECWORKS models need to be verified.
2.
Inspection data from 1995 and 1996 outages needs to be incorporated into
j
CHECWORKS models consistent with industry practice.
:
A
3.
The plant CHECWORKS models need to be updated and controls put in place
to document changes.
-
In addit 1Bn OPPD plans to conduct a follow-up self assessment of the
erosion / corrosion program following the next refueling outage to evaluate the
.
effectiveness of program enhancements.
.
$
During the implementation of our corrective actions OPPD will review and
j
incorporate, as appropriate, the following recommendations of the self
assessment team:
1.
Program Basis Document should be updated to:
a)
Eliminate duplication.
b)
Remove unnecessary detail.
c)
Make Program Basis Document and inspection procedure consistent.
2.
Program Basis Document should describe how susceptible systems are
dispositioned with respect to analysis.
.
-s
2
 
_
.
. . .- . ._.
.--
-
- - - - - - -
-
--
- - - - - - - -
*
.
:
.
l
-
3.
Program Basis Document should be updated to enhance communication
'
responsibilities between Operations and Maintenance regarding feedback
j
to Erosion / Corrosion Engineer.
4.
Guidance for documenting rationale for inspecting a specific location
should be provided.
5.
The inspection data form should be enhanced to address the following:
a)
Directions for the use of this form.
b)
Direction for what to do when measured minimum wall thickness is
inconsistent with nominal wall thickness.
c)
Direction on what to do when previous thickness is not available
and nominal wall thickness is not used.
d)
Clarify the step " List expansion of test sites not previously
evaluated."
-
6.
Typographical errors in the Program Basis Document should be corrected.
7.
Typographical errors and inconsistencies in the inspection procedure
should be corrected.
m
8.
The highest allowable value of maximum allowable stress (SE) should be
used to eliminate unnecessary conservatism and ensure consistency with
CHECWORKS.
9.
Program documents should be revised to proceduralize the following:
.
a)
Trending of inspection results.
b)
Qualitative evaluation of inspection data.
c)
Evaluation of data to ensure thinning is bounded.
10.
The acceptance criteria (Design Minimum Wall) being used for inspection
data evaluations should be reviewed to ensure that OPPD applicable code
requirements are met.
.
11.
The grid size on 6" components should be reduced to comply with
recommendations in NSAC 202L.
12.
A clear separation in Erosion / Corrosion Program documentation between
Flow Accelerated Corrosion (FAC) and other wall degradation mechanisms
(such as raw water corrosion) should be provided.
.
3
 
-
-
-
.
.. - - .
- _ -
.
-
-
.
- . .
. -..
'
.
,
i
.
.
13.
The guidance of NSAC 202L should be employed to address susceptible
,
piping that is not suitable for modeling.
14.
Complete CHECWORKS models of susceptible piping that is suitable for
modeling should be developed.
15.
The isometric-drawings identifying CHECWORKS component identifiers are
vital tools.
A set of these drawings should be placed in retrievable
storage and a second set should be used as a working copy.
16.
Sample expansions should not rely excessively on older inspections and
.
should aggressively seek to ensure thinning is bounded.
17.
Sample expansion on tee branches need to be performed on the train
'
,
containing the branch.
18.
The inspection data evaluation process should address how to handle
,
components with readings greater than nominal wall thickness.
-
'
-
19.
Any exceptions to the grid procedures should be noted on the layout
diagram provided in the outage summary notebook.
20.
The CHECWORKS Program should be used to perform inspection data
m
evaluation.
21.
Each inspection data package should include a printout showing the
inspection data matrix.
22.
Although it appears that informal communication does exist between
.
various departments and the Erosion / Corrosion Engineer the
Erosion / Corrosion Engineer should perform a review of emergent KdRs via
the Daily Emergent list.
Review of this list should give a heads up to
any developing system abnormalities.
23.
The closure review of configuration change and maintenance documents
should be strengthened to identify any issue of concern to the
Erosion / Corrosion Engineer.
'
l
24.
The Outage Scope Change / Addition Request form should be revised to
ensure that requests for deletions and additions to outages are properly
evaluated for Erosion /Cerrosion scope.
l
l
25.
Feedwater iron transport information should be added to the Seconday
Chemistry Monthly Summary Report.
;
26.
OPPD is a member of the CHECWORKS Users Group (CHUG). but is not an
4
 
1
*
.
i
d
active participant.
FAC personnel, especially the Erosion / Corrosion
'
Engineer, should attend the CHUG meetings. These meetings are held
twice a year and cover current Erosion / Corrosion technical issues as
well as a forum for discussing plant experiences.
27.
OPPD should participate in the CHUG Plant Experience Database to ensure
Erosion / Corrosion staff obtains future updates to this important
industry experience database.
_
28.
As a one nuclear unit utility, it is recommended that OPPD consider
joining with a group of other similar plants at other utilities to form
a peer group to share experience and peer assistance.
I
29.
A program to provide flow accelerated corrosion sensitivity training to
l
applicable plant staff beyond Erosion / Corrosion personnel should be
considered.
This will help to ensure that plant conditions that may
affect flow accelerated corrosion are communicated to the
Erosion / Corrosion Engineer and incorporated into the Program.
4
. .
30.
Use of resistant materials and systematic replacements should be
i
considered.
[
31.
The program should incorporate management involvement in important
program elements.
(Examples: CHECWORKS models, outage inspection scope.
!
Rg
outage close-out)
'
'
;
,
_
.
.
g
~~
5
l
 
_
. _ _ . _ _ . _ . _ . _ _ . . _ _ . _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . - - -. _ _._____._
i
,
.
!
!
l
Attachment 2
'
,
l
Fort Calhoun Station Root Cause
l
and Generic Implications Report
4
;
Fourth Stage Steam Extraction Line Rupture
i
CR 199700445 Revision 0
:
l
l
,!
!
!
l
-
1
i
!
;
it:.
:
'
-.
.
'
,
-
.
!
,
!
:
9
,
4
l
!
!
1
4
4
1
4
b
!
<
i
i
Dfj3
y.
-- --
--
-e
-
ee--
 
.
. . _ .
_
.
__
_
. _ _ . _ _ . _ _
_
_ . . _ _ _ . .
. _ _ _ _ _ . _ _
.
1
1
_
-
FORT CALHOUN STATION
,
.
ROOT CAUSE AND GENERIC IMPLICATIONS REPORT
FOURTH STAGE STEAM EXTRACTION LINE RUPTURE
CR 199700445
PRC RECOMMENDS
REVISION 0
APPROVAL'
i
SRG-97-026
MY 0 71997
,
'
t>HG MTG.-MINUTpgg
/
A. R. Patel, Lead Evaluator
,
Date
ShT7
,
[. R. Geschw nder, Evaluator
'
Date
-
W,
4~Y7
fK. G' asper, P/er Review Team Member - CR Owner
~
V ' ate
D '
"''
'
4]
e
1
G~/5/97
. L. Skiles, Peer Review Team Member
bate
..
.
R 4. M
**/s*/r'
R. L. JaIvorski, Peer Review Team Member
.
Date
/
f d'99
R. L. Phelps, Peer' Review Team Member
Date
b
Sf0Cf97
R. L. Andrews, Peer Review Team Member
' Date
%)&
s/sh7
M. T. Sweigart, PeeVReview Team Member - NSRG
D$te
#
dukb b
skk7
e su,
M. Kellams, supeNisor - HPES/RCA
Ddte
'
#
 
- . - - - . _ . - . _ _ - . - - - - - - _ _ . - . - . _ - - - _ _ . - . - . - _ . - . . . - . _ - . -
;
-
.
.
;
:
l
Attachment 3
'
!
-
!
Damage Assessme::t Report.
;
for the Break in the Eyiraction Steam Line
i
j
;
Revision 0, n'4ay 3,1997
-
.
l
i
:
!
!
!
!
.
I
!
;
:
!
!
4
:
!
I
i
l
i
{h
'%
i
!
;
..
.
.
1
:
i
!
i
!
!
!
;
:
i
-
b
.
.
1
i
1
$
i
!
!
!
i
i
f'
..
.
.
. - .
.
.
.
 
_._
*
-
.
.
.
DAMAGE ASSESSMENT REPORT
'
-
.
-
.
FOR THE BREAKIN THE
.
)
EXTRACTION STEAM LINE
..
Revision 0
l
.3
-
_.
May 3,1997
.
_
i
R. L. Phelps, P.E.
i
Manager - Station Engineering
M. R. Core, P.E.
-
,
,. Manager- System Engineering
 
- _ _ - _ _ _ - - - - _ - _ _ . _ _ - - . - - - _ . - - . -
_ - _ . - - . - - - - _ _ .
- - . -
i
.
.
.
i
!
;
.
!
-.
Attachment 4
'
4
i
!
!
Fort Calhoun Station
I
;
Erosion / Corrosion Program Assessment Report,
i
i
-
dated May 2,1997
-
;
l
4
!
i
!
l
i
i
:
1
3
i
}
l
l
i
A
..
.
I
f
,
!
!
.
l
1
/
i
i
_
-
l
<
4
i
!
;
i
l
1
I
4
1
1
?
-
!
k
W "b >
' D
. .
.
.
.
.
.
.
.
fp,
.
_ - - _ _ - _ _ , _ . . _ _
-
 
.
.
FORT CALHOUN STATION
EROSION / CORROSION PROGRAM
-
'
ASSESSMENT REPORT
.
*
.
.
_
.
l
m
-
Mt
> 7
L
P. Hopkins
b' @~
Craig Y%n[r (%P)
- .
C OS&
Ned Dietrich (DE&S)
e
Gus Undall"(Representinh EPRI)
,
CSM 4r Dw cl A. A.4k
.
David Smith '(DPC):
Approved:
%<1A
Me K.'GasperCo-TeamLeader/ 'Taylof
1
. o-Team Leader
.
tek
d
Q
i
ack L. Skiles
Sudesh K. Garrbhir
Co-Team Leader
* Sponsor
 
!
.
.
I
Attachment 5
'.
-
i
;
-
.
.
i
Fort Calhoun Station Self Assessment
j
Erosion / Corrosion Program Team Findings,
-
_
I
t
4
4
h
i
t
!
,
!
.
9
l
;
%.
;
I
-s
l
l
-.
.
!
:
!
-
,
l.
!
i
I
:
1
i
4
l
-
I
i'
i
;
i
i
!
,
t'
:
-
i
t
l
!
?
-
.
. . . -
.-
.- .
.
-
-
-
.. .
 
.
.
.
(~
j
R6 May L.
1997
Fort Calhoun Station SELF ASSESSMENT
Erosion / Corrosion Program Team Findings
Start Up Issues
!
Responsible
Corrective Actions
Completion:
Group
Date
1. Start Up Issues
A.
Upgrade the susceptibility evaluation to ensure
B. Lisowyj
A Restart FAC Susceptibility
05/08/97
(Should be
the following:
R. Aleksick
Review was performed to define
Complete
addressed prior
a)
Susceptible systems are covered.
D. Rollins
the scope of the FAC
Closure
to restart from
b)
Susceptible segments of susceptible systems
susceptible piping. As a
Memo
current Forced
are identified.
result the following
FC-0019-17
Outage.)
c)
Operators input is incorporated.
additional lines were added to
Section 2.0, F3
d)
Current susceptibility criteria are applied.
the program: Seal Steam
e)
Susceptible segments of susceptible systems
(entire system). S/G Blowdown
need to be addressed.
(suction / discharge of BD
transfer pumps). Condensate
Provide NRC the analysis and justification for
Recirculation (recirc.). Steam
not test.
Traps and Drains. Complete
any additional needed
inspections Pre Start-up.
Continue review to include
lines such as small bore.
Section 4.0, F1
B.
Review systems to ensure piping and components
R. Ruhge
Review MW0s. Mods. past
05/05/97
downstream of replaced components have been
R. Frakes
inspection data to determine
Complete
inspected to ensure industry experience has been
K. Hyde
inspected locat, ion. Perform
Closure
addressed. Document report.
an inspection not previously
Memo EOS-
Followup justification of why exclude in the past
completed. MWO 971649
SSE-97-065
to provide to NRC later.
(Complete)
Section 4.0. F2
C.
Component S-56 appears to have been installed
R. Jaworski
Inspect 5-56 and documentation
05/08/97
without the required reinforcing pad.
K. Woods
to determine thickness of
Complete
(Documentation is being pursued by OPPD personnel
component. Modify if
Closure
.
that may resolve this issue.) (Prior to Critical)
necessary.
Memo EOS-
ECN 97-161. CWO 97-037 (Comp..)
SSE-97-066
'
.
.
 
.
.
e
.
.)
R6 May '
1997
Fort Calhoun StoCion SELF ASSESSMENT
Erosion / Corrosion Program Team Findings
Start Up Issues
Responsible
Corrective Actions
Completion
,
'
Group
Date
Section 4.0. F3
D.
Packages from the 1996 Outage have'not been
D. Rollins
Review pkg 5-4. S-33. 0-26A.
05/08/97
l
independently reviewed (Examples: S-4. S-33.
A. Patel
S-38. D-84A. D-213 and other
Complete
D-26A. S-38. D-84A. D-213) as of 4/30/97. (Prior
pkgs if identified.
Closure
i
to Critical)
Memo
,
FC-0020-97 i
Section 4.0 F4
E.
Components displaying significant wear should be
N. Dietrich
Re-evaluate-components using
05/09/97
re-evaluated using industry standard techniques.
industry techniques. Identify
Complete
(Examples: S-73. S-74. S-66 and S-63)
needed inspections as needed.
Closure
Provide Technical discussion.
Memo
FC-0022-97
Sectior 4.0. F7
F.
Data or evaluations could not be found for the
D. Rollins
Locate documentation and
05/09/97
some 1996 inspection locations. (Examples: 5-57
R. Ruhge
include in database.
Complete
S-80. 5-92) (Prior to Critical)
Closure
Memo EOS-
SSE-97-068
+
i
*
i
l
'
.
.
 
.
.
. _ . _ . _ - _ . _ . . . _ . _ _ _ . . . . . _ . _ . . _ _ _ . _ _ _ _ _ _ _ _ . _ _ .
_
.
.i)
R6 May
'1997
b'
FortCalhounStationSELFASSESSMENT
Erosion / Corrosion Program Team Findings
!
Start Up Issues
,
'
Responsible
Corrective Actions
CompletionI
,
Group
Date'
!
Section 4.0. F9
G.
A review of high priority systems (Feedwater.
D. Rollins
Review the 6 hish priority
05/08/97
L
,
Steam Dump and Bypass. Blowdown. Extraction
B. Lisowyj
systems and identify points
Closure
Stean Condensate, and lleater Drains) should be
needing inspection. Complete
Memo
!
,
performed to ensure locations that industry
inspections or verify that
FC-0021-97 '
experience has shown to be potentially.
inspections have been
t
susceptible have been addressed.
performed. Complete necessary
:
Meeting on how we came up with these system and
repairs / replacements as
selection criteria.
necessary.
Inspection MWO's:
!
Complete:
i
971627(ES-3A). 971629(ES-2E).
I
971630(ES-2C). 971632(HD-3A.
>
38. 3C). 971649(11D-18).
971674(HD-IH) 971703(11D)
971666(SGB-2C)
Repair NWO's:
Complete:
I
971650(HD-3A). 971651(11D-38).
!
971652(llD-3C). 971655
;
.
t
i
!
B
'
t
!
.
t
.
.
.
s
___
._ _ - _ _ - - - - _ _
- -
 
_ _
m
..
_
._
_.
. _
.
.
*
i
FsrtCr,thouni.a$
1
R8 Hi
997
+
,
ion SELF ASSESSMENT
Erosion / Corrosion Program Team Findings
short Term / tong Term issues
Responsible
Corrective Actions
CceptetIJn
Group
Date
2.
Short 1erm issues
A.
Procedures should include more specific guidance on how
B. Lisowyj
Provide additional specific guidance
1998
(Should be
Outage inspection locations are chosen.
I
Procedure
for determining outage inspection
Refueling
addressed prior to
Group
locations in the Program Basis
Outage e
restart from the
Docunent (PD8) and other program
,
1998 Refueling
documents as needed.
Outage.)
Srction 1.0, F1
CID 970567/01
Szction 1.0, F2
s.
Measured wear determination process should incorporate the
8. Lisowyl
Incorporate industry practices for
1998
following industry practices:
Procedure
measured wear determinations in
Refueling
a)
(Jse of accepted practices to determine lifetime
Group
appropriate program documents as
Outage
comonent wear (circunferential band, moving
follows:
blanket, point to point)
a)
use of accepted practices to
b)
Clarify the use of engineering judgment relative to
determine lifetime component
wear determinations.
wear (circumferentist band,
c)
Process for incorporating measured wear into
moving blanket, point to point)
CHECWORKS models.
b)
Clarify the use of engineering
judgment relative to wear
determinations.
,
c)
Process for incorporating
measured wear into CHECWORKS
models.
CID 970567/02
Section 1.0, F3
C.
Sagte expansion process should be revised to align it with
8. Lisowyj
Revise the sagte expansion process to
1998
Industry standards. Specific changes include:
be consistent with industry standards
Refueling
a)
Clarify wording for smatt bore piping.
as follows:
Outage
b)
Add requirement for inspecting upstream of
a)
Clarify wording for small bore
expanders / expanding elbows,
piping,
c)
Clarify that pressure /teverature exemption only
b)
Add requirement for inspecting
applies to raw water systems.
d)
Clarify that expansion is to patattet co m onents in
upstream of expanders / expanding
elbows.
'
each train.
c)
Clarify that '
e)
Define the terms " component" and " highest wearing."
f)
Specify highest wearing co monents in the same
pressure / temperature exemtion
only applies to raw water
train.
systems,
d)
Clarify that expansion is to
parattel components in each
train.
e)
Define the terms "co monent" and
,
" highest wearing."
.
f)
Specify highest wearing
components in the same train.
CID 970567/03
:
.
-
. -
.
.
.
.
.
. .
.
.
.
.
.
. .
.
.
.
.
.
.
 
__
m
. _ _
.
.
t
R8 Mr
s e997
i
fort Csthounl
T,lon SELF ASSESSMENT
,
'
Erosion /Corrostw/>rogram Tsam Findings
Short Term /Long Term Issues
Responsible
Corrective Actions
Completion
Group
Date
-
S2ction ?.0, F1
D.
A docunent is needed to describe and control the
S. Lisowyj
Revise existing documents or deveicp
1998
Identification of susceptible systems.
'
Group
the process for the identification of
Outage
Procedure
new docments to describe and control
Refueling
,
'
susceptible systems.
,
CID 970567/04
,
Section 2.0, F2
E.
A docunent is needed to describe and control the evaluation
B. Lisowyj
Devel>ip or revise existing procedures
1998
.
of susceptible systems.
Procedure
to describe and control the evaluation
Refueling
Group
lofsusceptiblesystems.
Cutage
CID 970567/05
Section 3.0, F1
F.
Current plant CHECWORKS models need to be verified.
9. Lisowyj
Verify the current CHECWORKS modet.
1998
Refucting
CID 970567/06
Outage
Section 3.0* F2
G.
Inspection data from 1995 and 1996 outages needs to be
B. Lisowyj
Input inspection data from 1995 and
Prior to
Incorporated into CHECWORKS models consistent with industry
1996 outages into CHECWORKS n'.odel.
1998
practice.
Refueting
CID 970567/07
Outage
Stetton 3.0* F3
H.
The plant CHECWORKS models need to be updated and controls
B. Lisowyj
Ltpdate plant CHECWORKS model and
Prior to
1
put in place to doc m ent changes.
Procedure
revise existing documents or develop
1998
Croup
new docments to adninistratively
Refueling
contret revisions to CHECWORKS model.
Outage
Clu 970567/08
Srction 4.0* F5
3.
Docunentation should be provided when grid refinement or
B. Lisowyj
Revise existing docunents or create
1998
scanning is performed when significant erosion / corrosion is
Procedure
new documents to require documentation
Refueling
'
found.
Group
when grid refinement or scanning is
Outage
performed when significant
erosion / corrosion I,s found.
CID 970567/09
Section 4.0' F6
J.
The inepection data evaluation for components with PASS 2
B. Liscwyj
Revise inspection data evaluation for
1998
CHECWORKS analysis did not consider current predicted wear
components with PASS 2 CHECWORKS
Refueling
rates.
analysis to consider current predicted
Outage
wear rates.
.
CID 970567/10
.
.
.
 
__ - __-
.
. - - - - _ _ - _ - - _ - - - _ _ _ - - _ _ - _ - _ - _ _ _ - _ - _ _ - - - _ _ _ _ _ _ _ - _ - - - - _ _ - - - _ _ _ .
_ - _ - _ _ - _ _ _ - - _ . _ _ _ _ _ _ _ - _ _ _ . - - - _ _ - _ _ . _ - - _ . . _ _ _ . _ - - . - _ _ _ _ _ _ _ , - _ - -__
.
.
.
/'
,
E)
RS >.
I, i997
Fe:rt C:lhoun ..ttion SELF ASSESSMENT
Erosicn/ Corrosion Program Team Findings
Short Term /Long Term Issues
Responsible
Corrective Actions
Completioni
Croup
Onte
Section 4.0, F8
K.
The program does not appear to employ verification of many
8. Lisowyj
Revise doctanents or create new
1998
elements and a formal process for changing program elements
Procedure
documents to strmgthen adsinistrative
Refueling
does not appear to exist. (Examples: CHECWORKS model
Group
control of verifir.ation and changes to
Outage i
changes, Erosion / Corrosion Program General Information Table,
program elements (Examples:
Erosion / Corrosion Program Technical Data Revleu.)
CHECWORKS model changes,
Erosion / Corrosion Program Generat
Information Table, Erosion / Corrosion
Program Technical Data Review.)
CID 970567/11
3.
Long Term tstues
A.
A follow-up assessment should be performed following the next
J. Casper
Perform a format assessment after 1998
Post 1998
S ction 9.0, F1
refueling outage to evaluate the ef fectiveness of program
RF0 to evaluate the effectiveness of
Refueling
enhancements
program enhancements
Outage
CID 970567/12
I
.a
e
4
4
_ _ _ _ _ _ _ _ _ . . . _ . , _ _ _ _ _ . _ . _ _ _ _ _ . . . _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
-
. _.
.
. - -
. - _ _
_
- _.
.
.
l
Enclosure 3
-
Fort Calhoun Station
NRC Handout
i
,
t
;
i
;
.
4
 
, . . . . - . - . . - - . . - -
. . . . - - . . ~ . . _ .
. - . - . - - . . - . - . - - ~ . _ . - .
_ - . _ . - - . . .
).
a
. . .
.:
..
,
;
PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA
4
.
'
CONFERENCE WITH OMAHA PUBLIC POWER DISTRICT
,
:
July 21,1997
.
i
NRC REGION IV, ARLINGTON, TEXAS
,
4
1.
' INTRODUCTIONS / OPENING REMARKS - Ellis Merschoff, Regional Administrator
2.
ENFORCEMENT PROCESS Michael Vasquez, Enforcement Specialist
'
i'
3.
APPARENT VIOLATIONS & REGULATORY CONCERNS - Dwight Chamberlain,
j.
Deputy Director, Division of Reactor Safety
1
i
4.
LICENSEE PRESENTATION -
!
{
5.
BREAK (10-MINUTE NRC CAUCUS IF NECESSARY)
i
j-
6.
RESUMPTION OF CONFERENCE
7.
CLOSING REMARKS - LICENSEE
8.
CLOSING REMARKS - Ellis Merschoff, Regional Administrator
,
,,-,-,, _- ,
, . - -
. - . - , . - ,
, , , , _ . . . . , . - - .
. . , - .
.
-
 
-. .
.
.-
..
. - . -
.
-
_ . . .
. - - . .
.
- -_ .
..
,
.
,
.
f
i
;
'
,
[
APPARENT VIOLATION *
;
;
i
!
PREDECISIONAL ENFORCEMENT CONFERENCE
!.
!
.
OMAHA PUBLIC POWER DISTRICT
e
[ JULY 21,1997]
.
1.
i
* NOTE: THE APPARENT VIOLA TION DISCUSSED A T THIS PREDECISIONAL
ENFORCEMENT CONFERENCE IS SUBJECT TO FURTHER REVIEW AND MA Y BE REVISED
:
PRIOR TO ANY RESULTING ENFORCEMENT ACTION.
!
!
:
l
j
4
*
J
,v
 
.
-
_
.
.
-
e
f
.
e
APPARENT VIOLATION
1.
10 CFR 50.65(a)(1) states, in part, that each holder of a license to operate a nuclear plant
shall monitor the performance of structures, systems, or components, against licensee-
'
established goals, in a manner sufficient to provide reasonable assurance that such
j
structures, systems and components, as defined in paragraph (b), are capable of fulfilling
their intended functions. Such goals shall be established commensurate with safety and,
'
where practical, take into account industry-wide operating experience.
10 CFR 50.65(b) states, in part, that the scope of the monitoring program specified in
paragraph (a)(1) shallinclude safety related and nonsafety related structures, systems, and
components as follows: (2) Nonsafety related structures, systems, or components: (iii)
;
Whose failure could cause a reactor scram or actuation of a safety-related system,
l
10 CFR 50.65(a)(2) states, in part, that monitoring as specified in paragraph (a)(1) is not
required where it has been demonstrated that the performance or condition of a structure,
system or component is being effectively controlled through the performance of
appropriate preventive maintenance, such that the structure, system or component remains
,
capable of performing its intended function.
Contrary to the above, as of April 21, '.997, for certain nonsafety related structures within
the scope of this rule, the licensee had neither monitored the performance of these
structures against licensee-established goals, nor demonstrated that the performance or
condition of these structures was being effectively controlled through appropriate
'
preventive maintenance such that the structures remained capable of performing their
intended functions. Specifically, the large radius piping elbows of the fourth stage
extraction steam system, sixth stage extraction steam system piping and other piping in
the heater drains system were neither monitored nor effectively controllad through
j
preventive maintenance such that these piping locations remained capable of performing
i
their intended function. This was evidenced by: 1) the second downstream large radius
piping elbow in the fourth stage extraction steam system failed catastrophically on April
21,1997, resulting in a plant transient; and 2) the following piping structures were
subsequently determined to be below minimum wall thickness: a) the furthest downstream
large radius piping elbow in the fourth stage extraction steam system line (S-32); b) a sixth
stage extraction steam system " pup" piece (S-54); and c) three parallel lines in the heater
drains system (D-95).
THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y
BE REVISED
}}

Latest revision as of 16:19, 24 May 2025

Summarizes 970721 Predecisional Enforcement Conference in Arlington,Tx Re Apparent Violation Identified in Insp Rept 50-285/97-09.Licensee Presented Summary of Causes & Corrective Actions.Attendance List & NRC Handout Encl
ML20149L208
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 07/25/1997
From: Howell A
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To: Gambhir S
OMAHA PUBLIC POWER DISTRICT
References
50-285-97-09, 50-285-97-9, EA-97-280, NUDOCS 9707310192
Download: ML20149L208 (111)
v • d • e


See also: IR 05000285/1997009

Text

.

.

..

._

.

'

i

e'

4

} 'S

p" "' 4 t

UNITED ST ATES

y

l

  1. $

%';

NUCLEAR REGULATORY COMMISSION

-

. . . ~ , ,

~' j i

REGloN iv

i

!

N.

'

C

e

i

S

f

611 RY AN PL AZ A ORIVE. SUIT E 400

j

o....*T

t

~

AR LINGTON, TEXAS 76011 8064

-

l

,

July 25, 1997

i

1

EA No.97-280

l

S. K. Gambhir, Division Manager

Production Engineering

Omaha Public Power District

Fort Calhoun Station FC-2 4 Adm.

P.O. Box 399

Hwy. 75 - North of Fort Calhoun

Fort Calhoun, Nebraska 68023-0399

Dear Mr. Gambhir:

SUBJECT:

PREDECISIONAL ENFORCEMENT CONFERENCE SUMMARY

On July 21,1997 representatives of Omaha Public Power District met with NRC personnel

in the Region IV office located in Arlington, Texas to discuss the apparent violation

identified 'a NRC Inspection Report Number 50-285/97-09. The conference was held at

the request of Region IV.-

The licensee presented a summary of the causes for the apparent violation and their

corrective actions.

The attendance list, NRC handout, and the licensee's presentation are encloseo to this

summary. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of -

this summary and its enclosures will be placed in the NRC Public Document Room.

Sincerely,

b

h Arth r T. Howell 111, Director

Division of Reactor Safety

,

i

Enclosures:

1. Attendance List

2. Licensee Presentation

(

k

3. NRC Handout

i

01OObb

Docket No.: 50-285

License No.: DPR-40

I b! !! h.h, !

9707310192 970725

gDR

ADOCK 05000285

PDR

~

.

J

.

$

.

Omaha Public Power District

-2-

cc w/ enclosures:

James W. Tills, Manager

Nuclear Licensing

Omaha Public Power District

Fort Calhoun Station FC-2-4 Adm.

P.O. Box 399

Hwy. 75 - North of Fort Calhoun

Fort Calhoun, Nebraska 68023-0399

James W. Chase, Manager

Fort Calhoun Station

P.O. Box 399

Fort Calhoun, Nebraska 68023

Perry D. Robinson, Esq.

Winston & Strawn

1400 L. Street, N.W.

Washington, D.C. 20005-3502

Chairman

Washington County Board of Supervisors

Blair, Nebraska 68008

Cheryl Rogers, LLRW Program Manager

Environmental Protection Section

Nebraska Department of Health

301 Centennial Mall, South

P.O. Box 95007

Lincoln, Nebraska 68509-5007

-

.

Omaha Public Power District

-3-

l

E-Mail report to T. Boyce (THB)

E-Mail report to NRR Event Tracking System (IPAS)

E-Mail report to Document Control Desk (DOCDESK)

bec to DCD (IE01).

bec distrib. by RIV:

'

Regional Administrator

DRS-PSB

DRP Director

MIS Systern

Branch Chief (DRP/B)

RIV File

Project Engineer (DRP/B)

Branch Chief (DRP/TSS)

Resident inspector

DOCUMENT NAME: R:\\_FC\\FCSUM.JL

To receive copy of document, indicate in box: "C"

Copy without enclosures

"E" = Copy with enclosures "N" = No copy

SRA:DRS

,l E

D:DRPA g

D:DRS [r),Y/, E

l

l

JFShackelfordJ

TPG W M

ATHo % I)I

s

07d_f97

l,

074fS7

1

07/25/97 "

OFFICIAL RECORD COPY

n

Omaha Public Power District

-3-

,

.4

E-Mail report to T Boyce (THB)

E-Mail report to NRR Event Tracking System (IPAS)

E-Mail report to Document Control Desk (DOCDESK)

bec to DCD W

. bec distrib. by RIV:

~ Regional Administrator

DRS-PSB

DRP Director

MIS System

Branch Chief (DRP/B)

RIV File

Project Engineer (DRP/B)

Branch Chief (DRP/TSS)

Resident inspector

.

DOCUMENT NAME: R:\\_FC\\FCSUM.JL

To receive copy of document, indicate in box: "C"

Copy without enclosures "E" = Copy with enclosures "N" = No copy

SRA:DRS

E

D:DRPA, ,gg D:DRS f,Y/, E

,

JFShackelfor4

TPG W M

ATHoMMI

07F/97

[,

074Ff97

1

07/25/97 "

OFFICIAL RECORD CO')Y

.

-

-

-

._ _

.

.-

.

..

._

. . .

1

J

, .

'

,

I

.

I

ENCLOSURE 1

1

Fort Calb.aun Station

,

'

Predecisional Enforcement Conference Attendance List

!

l

l

l

i

._

_ - - _ _ _

_

-

_

__

.

. _ _ _ _ _ . _ _

_ _ . _

'

.

'

'

PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE

j

LICENSEE / FACILITY

Omaha Public Power District

Fort Calhoun Station

DATE/ TIME

July 21. 1997. 10:30 a.m.

CONFERENCE LOCATION

Region IV. Training Conference Room

.

Arlington. TX

{

EA NUMBER

EA 97-280

NRC REPRESENTATIVES

,

)

NAME (PLEASE PRINT)

ORGANIZATION

TITLE

W l l? H 1 C M (4 fn 6 L f L % H

HK&

0%V1y

ofpic sa

o

Fut3

mMscHoW

H%

RMIoans

(40knmqmfoe

b

94WP2foEd)

/U M

9 . M M 7tt h

sch CGJ NL

%k Owas

uec

A4k4d' Ihssw2

Nac 2iV

G4~

s %.u tE

~/

v

DAlf A. Ecujweg

NRc.

RS

CHtEF, /1&idre4gcc ?gM::s

/f.

. h/fGW

&C

RT

/ ES/GA/A L {cc&dfEL

V' Ni6 L

NRC/N2R

D(of

6.k@ W

l

Pd.t h

hW

ec m

'

W :llta , O. SL s. ,

W R c- Ki]L

R n <i ct,i e-f/DRP

Page _ of _

.-

-

.

.-

.

-.- .

. - . . -.

-.

-

"

'

.

,

!-

PREDECISIONAL ENFORCEMENT CONFERENCE ATTENDANCE

,

l

LICENSEE / FACILITY

Omaha Public Power District

'

Fort Calhoun Station

i

DATE/ TIME

July 21, 1997. 10:30 a.m.

!

CONFERENCE LOCATION

Region IV. Training Conference Room

Arlington. TX

EA NUMBER

EA 97-280

s

LICENSEE REPRESENTATIVES

L

NAME (PLEASE PRINT)

ORGANIZATION

TITLE

%nta

c 4re<

c> f90

vp-acu su

'

g u g ,., e x. 6 a n n u

oppo

Di^^ M " P "

E g in"~'1

'

04 o}M nD~3

M orl

l

&& X/dae, war

G46,0

m aa+ w cewe m s

1

Jemes W Y;7h

0/Yb

% a f a - t % d u , G a d ,9

i

don ld. u e m

6fPC

%

% ROW NY-

y ,,,e

Hwchu hou

l

Wenn<u s.Dow A

oPPD

p) _ . mn l. gLb.

"

doseph KGaspe<

om

  • "M*"

.

l

k,b, kWCKGWJ

ONQ

Atacerette Asussinenb

/

VW/J/C*V N/J W

-

svue seaues aose

DAU10 l .

OLL /Wh

D hh0

' 0'd6Mm Even - EMiut/tonoved

l

,

Page _ of _

1

l

.

ENCLOSURE 2

Fort Calhoun Station

Licensee Presentation

l

j-

.

.

,

!

l

OMAHA

!

PUBLIC POWER .

!lo

DISTRICT

i

i

!

Fort Calhoun Station

Main':enance Rule

i

!

Prec ecisional Enforcement Conference

4

f

l

July 21,1997

!

!

!

)

llO

l

1

l

_

_

_ _

_ _

_ _-

_______

e

-

e

... ... ;.

.

-A.ng.

  • 1ggg g

-

--

,

Opening Remarks

Introductions

e

Gary Gates

e

2

- _

- - .

-

- - -

-

- . - - - - -

- -

.

.

$

l

l

l

.

<

lO

4

J

-

. , ,,, n

.

. . .

. , , , . . . . . . . . . _ . _ _ . . , . .

..

-

..

I

!

l

.i

4

l

1

i

)

.

!

'

4

l

Agenda

~

,

f

-

~

O

Sudesh Gam ahir

O

3

._.

-.

>

.

_

Agenda

- _ .

o Operational Overview

Ross Ric enoure

o Maintenance Rule

Jim Tills / John Johnson / Ken Dowdy

-

o Assessments and Corrective Action

O

Update

Joe Gas aer

o OE Program Assessment

Dick Andrews

o Summary and OPPD Persaective

Sudesh Gambhir

o Closing Remarks

O

Gary Gates

4

_

_ _

_ _

- _ _ - - _ _ _

e

1

-

i

,h

.

.

i o

,

ii

TYfMd7W 'FV.]

.

.,

i

_ _ . . . . .. . . , , . ..

!

!

!

I

i

)

!

l

l

Operational Overview

,

O

,

a

$

!

Ross Ridenoure

i

l'

1

l

'

,

!

1

l

i

!

l

!

i

!!O

1

i

'

1

5

i

!

- -

-

. - - .

-_

_ _

. - _ _ - _ _ - - - _ _ - - . - - - . -

)

.

Operational Overview

.

.

O

-

.

1

.,-sesvp pg

tg,g. gy p ,, _

, -

~

o Event Overview.

o Primary Plant Impact.

.

o Secondary Plant Imaact.

O

o Operational Safety Significance.

.

O

6

-

________-_-_

_-

,

,

l Overview of the Rupture.

!O

i

.

l

xv;n~;._

l

i o Major steam rupture in 4th stage

I

extraction steam line.

i

i

l o Reactor tripped within 19 seconds.

! O Crew entered emergency procedures

and stabilized the plant quickly.

o Crew responded in a decisive, timely

manner.

o NOUE declared and ERO activated.

O

7

l

Primary Plant Impact

-

!

e

-

.. i f " *h Z;u..._ .

"

j o Prior to the trip the steam rupture

!

produced no changes in reactor power,

steam generator pressure, or other

l

primary parameters.

.

I

i

' o Normal Post-Trip response observec. 4

.

o Operatinc crew made a cecision to

emergency borate the reactor.

o No significant challenge to the operating

crew.

e

8

.

.

. _ .

.

_

_ _ _

_ _ _ _ _ _ _ _

'e

j. Sec6ndary Plant Impact

.

!O

1

_

~w..;gappymast w ? .

-

,

i

l o Damage Assessment- Crew

l

Investigation

l

!

!

-

I

One MCC was c e-energizec

.

!'O

,

i

Some piaing in the immediate area o"the

!

rupture was bent / twisted.

.

!

l

l

!

!

Wetting of Equipment.

!

i

}

t

!

!

Steam Volume contained in the Turbine

j

Building.

!O

i

!

9

.

___ __

___ ______

!

.

l

Sec6ndary Plant Impact

-

,

.

!

4

e

'

'd

.' - .: '

r

.' ; 4 + , ,. .

o Fire suppression actuated in the

.

Turbine Building basement anc

mezzanine.

.

Intermittent ground alarms on a 480V bus

and DC bus #1.

O

No adverse plant effects were observed

due to t1ese intermittent grounds.

0

10

-.

.

. .

.

-_

l

Qperational Safety

.

tO

j

Significance

-

.

l

- . 22:= r .x

! o Event effectively mitigatec by the

j

o aerating crew.

[

Minimal effect on the arimary plant.

-

l

Minimal effect on t7e secondary plant.

!lO

Minimal reduction in Fire Protection System

l

capability.

i

,

l o Not a significant operating challenge to

1

i

the crew.

l

l

l

i

l

lO

!

11

i

.-

__

_

_

_

_

-

-

---_- _ -

_

.

.

.

e

~~~ru :. . .

l

~

Maintenance Rule

e

Jim Til s / John Johnson /

-

Ken Dowdy

9

12

'

_

---

-_

-

_ _ _ -

--

-_--

.

Maintenance Rule

-

.

Discussion

!

y__ .

o Scoping of Systems Under the

Maintenance Rule.

Goal Setting / Performance Criteria

,

requirements

-

o Performance Monitoring of the

O Extraction Steam System.

EC program monitoring effectiveness.

o Use of Industry Operating Experience.

Program Development.

Post-Event Corrective Actions.

o Imalementation of Maintenance Rule

Requirements.

O

13

-.

-

_ _

_ _

.

. . _ . . .

.

.-

.

-

_

_

_

__

_

_.

Monitoring Under the

.

.

Maintenance Rule

ygn;=_

.

o Sco3ing of Extraction Steam.

Included within scoae of program as part of

t1e Main Feedwater System.

- No failures of Extraction Steam had caused a

plant trip at FCS.

O

- Industry review indicated that Extraction Steam

piping could potentially cause a plant trip.

Classified as Nonrisk Signi"icant

- Based on FCS PRA.

- NUMARC guidance.

Reviewed and Aaproved by an FCS Expert

Panel.

O

14

.

. _ _ -

.

_-_

_

_

_ -

- .

,

!

!-

Monitoring Under the

.

lU Maintenance Rule(cont)

- - _,_

y

, : , , _ u-

_

,.-.c

! o Extraction Steam SSCs were monitorec

l

using Plant Level Performance Criteria.

l

l

NUMARC guidance provides that Plant

l

Level Criteria are ap aro ariate.

~

,

i

Plant Level Performance Criteria

lO

established were:

!

!

- No Plant Trip due to MPFF.

!

l

- No Unplanned Capability Loss due to MPFF.

!

- No Safety System Actuations due to MPFF.

System Level Performance Criteria:

- System must be available (100%) when required

for power operation.

g Reviewec and A3provec by Ex3ert

Panel.

15

_

_.

_ _

_ _ _ _ _ _ _

- -

Monitoring Under the

-

.

Maintenance Rule / cont >1

\\

1 , f ". r " : * i . i -

o Evaluated against Plant and System

Level Performance Criteria.

Review of SSC failures and maintenance

history was performed from 7/1/92 to

6/30/95.

System monitoring ongoing since 7/1/95. O

No failures prior to the rupture were

identified that exceedec the Performance

Criteria.

o System placec in Category (aX2} of the

rule.

o Reviewed and Approvec by Expert

Panel.

O

16

-.

.

.

.

. , . . _ . . ~ . .

!-

l-

Monitoring Under the

.

l Maintenance Rule (cont)

a

=u r_ .

o Piping considerec effectively controllec

l

by Erosion Corrosion Program.

!

l

l

Review incicatec wel-developed inspection

j

prog ram.

O

Proactive replacement of piping prior to

reaching minimum wall thickness.

O

17

_ _.-.__

_____

i

-

-

l Status of Extraction Steam .

l Piping After April 21,1997

l

.. _ r: m e t _

.

i

i o Actions taken as a result of the

!

Extraction Steam Line Break.

l

Piant Level and Main Feedwa":er System

'

Performance Criteria exceeded.

Cause Determination performed based on

RCA and Failure Analyses.

g

- Identified Failure as MPFF.

- Recommended all FAC-susceptible piping be

placed in Category (a)(1).

o Findings and Recommendations

Reviewed and Approved by Expert

Panel.

o Additional Evaluation.

All plant system piping reviewed.

$

18

.

.

-

.

.

.

.

.

l Summary of Maintenance

o

l

Rule Compliance

-__

,

o Industry operating experience was

properly taken into account. ('" 9.3.3}

4

i

o The extraction steam system classifiec

!

as "nonrisk .significant". ('" 9.3.2}

llO

Guidance. allows monitoring using plant

level criteria.

- No automatic reactor scrams.

- No unplanned safety system actuations.

- No unplanned capability loss factor.

System level aerformance criteria was also

established.

- Requires 100% availability during power

operation.

O

19

.

Summary of Maintenance.

.

Rule Compliance

.

.

. 1 enre.1. .

,,

o The failure history on extraction steam

was reviewec for a four year period

prior to July 1996. {'" 9.3.3}

.

No problems were detected involving

extraction steam (no through wal leaks ag

no alping below minimum wall).

o Based on failure history, extraction

steam did not require saecific goal

setting and was correctly monitored in

accordance with 10CFR50.65(aX2}.

{'"9.3.4)

e

20

__

.

.

_

_

_ _

_ _ .

_ _

_ -

.

.

l Summary of Maintenance

.

O

l

Rule Compliance

l

. : :n;=rw

i o Extraction steam was monitored by the

j

EC program. At the time of Maintenance

l

Rule implementation the EC program

l

was judged to be ef ective.

.

!

Proactive replacement of piping

iO

mponents.

!

l

'

Effective in meeting performance criteria

and preventing repetitive MPFF.

Highly susce atible extraction steam

components extensively monitored (70%

sites inspected).

Multia e

ssessments indicated t7e

O

3rogram was an Industry leac er.

21

. _ - .

--___--____

_

_

!

-

l Summary of Maintenance.

l

Rule Compliance

i

.

___

-

.

.

.-,... ...

..

.

.

,,, . .. .. .

.

l o The April 21,1997 failure caused both

plant and system level performance

'

!

criteria to be exceeded.

!

!

-

!

!

l

Failure was the initial MPFF.

1

l

l

No extraction steam MPFF had occurrec

'

prior to Maintenance Rule implementation.

This is the first known large radius swee a

failure in the industry.

O

22

--

-

-

__

__ _______

_ _ _

!

-

l Summary of Maintenance.

lU

Rule Compliance

.

_

l o Based on the failure the following

l

actions were taken:

1

!

A cause determination was performed.

FAC susceptible piping was placed in

l

category (a)(1).

iO

i

Goals were established.

l

Industry experience was again used during

!

both the goal setting phase and cause

determination. (C 9.4.1 and 9.4.4)

,

!

!

o A determination was made on whether

j

other piping systems within the staae of

l

the Rule were being effectively

j o monitored by the EC program.

!

23

,

1

i

-

.

.

- -

--

-

.

_ . _

- - _

_

.

.

l

Conclusion

.-

e

~x t

_

l o While a Significant Event, the Failure Is

Not, in Itself, a Violation.

!

!

!

' o The Rule Worked As Intended.

~

l

l

l o Extensive Corrective Actions Were

l

Taken.

l

o Industry Experience was utilized per

NUMARC guidance.

o No Violation of the Maintenance Rule

Occurred.

O

24

-

.

_

.

.

a

aru.--n2musa..

s

s -m eu

aesu-asma>sms-ma-ea,nsn2.----.

.u.**-,sa,+ona.

.x.*..

so.<=a=a-

    • .um-*-~-.

mm+*-~~a-=sman--=-~,-*-

+ , -

a~~.s-+a----a

-

i

.

!.

f

=

lO

.

i

'

=_ .n,

!

!

!

l

Assessments and

i

.

l

Corrective Action

l0

Update

i

i

!

Joe Gasper

O

25

-

.

-

_

_ - - - _ _ _ _ - _ _ _ -

.

'

l

Assessments

Objective

.

- , - _

-

__.........................,.,.s......

..

l

!

!

!

!

To ferret out the root cause

and any other program

,

deficiencies or problems.

4

9

26

.

.

.

-

. - -

-

-

- - _ _ _ _ - . - _ _ . _ - _ . _ _ -

I

-

i

.

i

Assessments

-

.

l0

!

l

!

-

. wico, m ,,,, .r. c

.

<..,,n,

.m

.-.,;.

$

i o Initial Assessments

i

-

'

i

!

i

~

i o Additional Assessments

!lO

1

o Corrective Ac: ions

!

4

}

!

,

i

i

I

!

27

.

_.

.

. _ - - . _ _ _ _

-

-

-

_ -

.

.

Corrective

-

Assessments

Actions & G

I

Results

.n ,

- ,

-...; w v.w w.wa - . - 2* gym - - -

er, . . . .

l

+ Lessons

W_=.,

5

-

ong Term

l

+ Corrective

lW-

'

28

. . - -

.

...

-

-

. .

.

.. .

- . .

. - -

-

.. ...

_ _ - _ _ _ _ _

_ _ _ _

l

.

-

LAdditionalInspections and

lU

Replacements

l

,=

-

l o 23 Sites Inspected

o Replaced because of FAC

l

4th stage sweeps.

6th stage 18 "- 45o elbow - Conservatively

,

l

Replaced - Would Have Reached Design

l O

Minimum Wall in 3 0 3erating Cycles.

o Replaced for other reasons

.

l

6th stage 18"- 45o elbow and pipe - Weld

)

fit-up aroblem aroduced locally high

'

turbulence.

Heater Drain - Three Identical 3 " Pipes

[

downstream of orifices - Visual examination

{

of pipes found no localized wear.

!O

In lusions orlaminations may have causec

l

erroneous UT indications.

29

I

. .

.

-

-

-

. .

.-

Altrrn Ccrporttiin

Technical Report N . 97152-TR-01

'

Revision 0

i

O

R

=36

0lS;

-

kHgAsq;

A

CI q%

At4

!

$3

R'

$

Q

3

  • /

199

g

IN3

.

e.

e

,6

N

SS

y3

-

43

IN

.

gC

R

0474

f

IN

1

P(q 04

>92

gwl,{ ,fj!;

2

%s,

s

-

Ett 6g

ER

6 D

0

g

I4

6-

IN

E(

%$, $1

$.

No

0A74

4

,,,

9

y

N

4-

e,.

%

NOTE: % WALL LOSS CALCULATED FROM too. AT INSPECTION DATE INDICATED.

FORT CALHOUN 4th STAGE EXTRACTION

STEAM LINE TO F.W. HEATERS

,,g;;yggL,

Geometry & Inspection / Replacement

Sununary of 4th Stage Extraction

.

-_

-

_

__

._ .

_

i*

-

l

Replaced Components

.

o

l

i

.

.

.

+ :aguansum;;

-

.

.,..,

.a.

...

. p .

. . .

.

. . . ,

i o Conclusion

,!

.

l

Two sweeps required replacement:

.

l

- The ruptured sweep elbow and

.

l

- The 10" sweep due to FAC (below rninimum

l

wall).

!O

l

Two components showing FAC were

,

l

conservatively replaced (above minimum

l

wall}.

l

l

Four com 3onents replaced for other

!

reasons.

i

!

!

l

!

l

30

i

.

Update to Assessments

-

.:: : w n

..

o Failure Analysis

o FAC Code Verification

-

o FAC Program Implementation Review

Additional Information Concerning

Replaced Com aonents.

O

'

31

_ - _ _ - _ _ _ _ _ _ _

_ _ _ . . _ _ _ _ _ _ - _ _ - _

i-

.

.

Failure Analysis

.

l

Completed

-

. - -

!

l FPI and Altran Concluded:

l

l

i

j

The Root Cause was Flow Accelerated

,

l

Corrosion.

l

llO

!

There is Evidence of High Velocity Water

DropletImpingement.

i

Complex hydraulic arofile aroduced large

variation of oxide accumulation and

damage characteristics in failed sweep

elbow.

O

32

.

.,.__wa__

4

-*u,.J.

4-m.-am-e-.4.--ie4m

e eis - 4

eal m..a._sm,4u2hh=4e

e.enN J

e

e.Msa-J A eM Aa.-a4-4

-

-,e.cet

-

""ada4.,dezpa.h 4, AA 4 p A se .mm 4-

-

-A4,.e,ide.;W.M_.

@

Water Droplet

-

l

Impingement Craters

I

c.

eny g g 4 g

j;g.y...

.

.

.,

-

,

I

i

.

s..

e

c.

, w . wi . . ., ;.,

m 3 o r- - - -- ur p.r. ,. . . . .

.

I

1

j

> T.

.g

r.

. . y-p ,m m . .

w.a+

a

1

!

Y

QY*.Y.Ti T

~

.

P M. ,. a,e. ,9 P:% . , M -$' M ? G i d g

w.P7p..

u

i

,

,. ., . n. ,,, .- ..m e

-

.

.

..

.

h,$$fy;

_

..

s--

sa

s.

w

L

-

f. v.T"n

m% .-s. p;.

y3

. , . -g ; .

s e -. ' -

-,-T

ky

n . ny -

n

,

..

- a s r:w~~. .

-

-

-

gg

. . . ? , '.E;,i kk ... : C%: i

~'

-

    • t**u

!.

g <&

', NW; ' "" $

.

y

.

4

+

,,g,,..

. .g

,

s.n..?

+5.

.

%

~O Q L

-

^^ ,

~. .

W .,,

W _ y * ,^- ##. - .,~,g .,; !W:ft

j,

_,

.r

. . . . , .

Ey

' [M Utsc "w a p . ,

' t

1 ~ ' "'s

j - ",; * g; .*

W-

g

f**

.. . .**q

e .j

a

v.

-*

r-

,=

it%

., '.7^4. . *i '

'6 *;L*%.

,T

T., T T. . -

-

N,

g .. *J

f=j i * '

.

1

+4

[

, _

g.

y 's [-

h. w.*,j .,

~#

'58

  • '

-J2

t

..

- :

-

..

-

.

.ma

-

^ ^

,

.

-

py-

n g-

r; 4

.

2.

-

rj

. 1

. ~. s

.

s.

.g

.

ej

-

. _..

g ,- e % , gj- 4 _ % . %*a ,

n d w-

m, , v.m

s_

.

m

,

1

~;E '* V-

=

.

%

,

y ,a f . . i

W..' . t

.

- - - *

^T*

$ i.

  • k

Re

fy

&,.

-

j

T..

y

s

-

y

. , ' . . . - + %< 4 * * -

T+ *..

..

s -

.. .~ ..

, p ..

g

.

)

.

,

.

-4, I j f

. : -- -

,]

fC,

y

f .** d *

..i.

T-

. - 'y

  • "

<*-;

~

--+

-.

.

$ :,:, . =

h. " ..

* *

'. t

'

, ; ', +

,

,

~

y

..

^*

  1. '
  • _A

' J '-

f

. .

,-

_,

].t

. . -n

a.m

-

.*-

-

-a .

...

-g.~

- t

(_,,

-

'

i*

U od*J

~

.

7

, '

M

' %[ % g eWT .

.

':- u x .

  • - -

.1 x r

---

urt .

-

y

y

.%

O

33

- -

.. -

.

.

_

_ _ _ _

.

_ _ _ _ - _ - -

'*

i

.

l.

FAC Program

.

io

,

!

s u&2.?ttw*m :: . ,.

s

o Code Verification

!

!

l

CHECWORKS Modeling

'

l

1

-

l

BRT-Cicero Mode ing

O Program Implementation Review

!

l

Use of Plant 0 3erating Experience

i

i

!

Site Selection Methodology

l

Use of Industry Operating Experience

i

!

i O

i

i

34

.

.-

.

_

.

..

. ..

.

-

Altran C:rporrti:n

Technical Report

. 97152-TR-01

,

O

'

g5 N

%,

-

-

0t 2$>,

N

!\\,

IN

khgk

'#

b4IA

N

s

Ng

'

y

1

<

7

IN

-.

o.

r#

199

4

N

{93

I((,,@.

ing

.

,<gf

. ce2

"4

%$,$,?,hb

lt@

'

&?**

*"44

  • $$,3

9

'

,

,

%Qpp

"

hO

Cl

?.

g

>

o

2

9a7

E

1

OMx

4

NOTE: % WALL LOSS CALCULATED FROM t

AT INSPECTION DATE INDICATED.

FORT CALHOUN 4th STAGE EXTRACTION

,

,

STEAM LINE TO F.W. HEATERS

,,ggg g g ,,

j

Geometry & Inspection / Replacement

Summary of 4th Stage Extraction

.

.

-

-

-

-

.

..

l CHECWORKS Modeling.

!o

i

I

~

'

. < _

.

.. _ .,

!

o Update and Validate CHECWORKS

l

using 95,93 and 97 inspection data.

j

Parallel trains modeled.

l

Model Verified.

-

j

- Modeling errors corrected.

I

- Irispection data matrices entered for

lO

components.

!

! o Model updated with all data available

j

before 1993 outage.

l o Line correction factor.

Value reflects accuracy of predicted wear

!

rate.

Acceptable range between 0.5 and 2.5.

$ age

2

4

6

Factor

2.368

0.370

0.695

35

.

9

CHECWORKS Modeling.

.

e

(Cont)

,

yygny.;

o Measured versus Predicted Wear Plot

shows half of data points outside +/-

50% range.

.

O Conclusions:

e

A good corre ation is not established

between measured and aredicted wear.

CHECWORKS should not be used to

determine the wear status of components

that have not been inspected.

e

36

.-

.

.

-

-

.

$

!

BRT-CICERO Modeling

.

lO

.

. ; .2i"'Q

l o Code used by EDF.

!

l o Fourth stage modeled.

o Results

i

~

j

Predicted wear at end of Cycle 16 of 0.429

!

inches.

iO

i

Prediction ~is conservative to observed wall

loss of 0.057 to 0.151 inches depending on

,

l

actual initial wall thickness and a failure

thickness of 0.050 inches.

o Differences between Cicero and

CHECWORKS.

Uses component vs line correction factors.

Calculates individual component prediction

O

uncertainty.

37

.

Use of Plant Operating

-

e

Experience

mem_

o RCA Contributing Cause: Incomplete

.

utilization of plant operating experience

Additional replacement data identified

.

Now classed as a root cause

g

o Additional corrective action

Research, maintenance and configuration

changes done before program

implementation in 1988

,

O

38

_

-

-

_ _

_

_ _ _ _ _ _ _ _

l

Site Selection Criteria

.

io

!

,

!

'

r

.,um m ,x+.n

-

l o Applied NSAC 2 2L Guidance using

!

data available prior to 1993 outage

l

Inspect Highest Wearing

l

Components.

-

,

!

Shortest Relative Remaining Service Life.

l O

One Component from each parallel train.

i

j

Immediately downstream of control valve

l

and orifices.

i

l

One com aonent in each two phase line of

!

3iping.

Industry Experience.

'

Plant Experience.

Replacec components and components

within two pipe diameters of replaced

O

components.

Unusual geometries.

39

.

-

_ _ _ __ _ ___ _ _ _ _

Use df Industry Operatin(

Experience

..=~r u-

o Research of CHUG data base identifiec

no additional information and no

information relative to sweeps.

.

o Inspections basec on industry operatir$

experience.

70% of Extraction Steam sites inspected

before 1997 forced outage.

48% of the Heater Drain sites insaectec.

38% of tie Circulating Water sites

inspected.

35% of the Condensate sites inspected.

O

40

.

.

.

\\

_ _ - _ _ _ _ - _

.

9

FnC Implementation

.

.

l

Review Conclusions

i

~ . r., ., m, ye.

l o Industry experience effectively used.

!

}

!

o Plant exaerience not effectively usec.

$ CHECWORKS shou d not 3e used to

predict wear for components that have

not been inspected in the 4th stage line.

o Unique flow conditions may exist in t1e

4th stage line.

O

41

-

_

-

._

.

Current Corrective Actions-

t

e

i

Status

4

~

'

,.

-

.._m

-

,_...s...

o Conduct additional inspections to

develop PASS 2 mode s for

CHECWORKS.

o Evaluate on-line radiography for small

bore piping.

e

o Evaluate replacing high wear piping with

wear resistant piping.

o Evaluate additional moisture traps on

extraction steam piping to reduce wear.

o Provice additional training.

e

42

.

_

.

.

_

.

__

_ _

_ _ _ _ _ _ _ _ ___ _ ,

li Current Corrective Actions

!O

l

Status

.

.

., ,. .l.,(l ? ~

'

'#

o Work with EPRI to share experiences

l

with the industry.

l

l

i

~

l

l o Work with EPRI to improve modeling for

l O large radius sweeps.

i

l

!

! o Work with EPRI to better understanc

j

effects of oxygen concentration on

j

secondary systems.

!

!!O

!

!

43

. --

. . _ .

- ---

-_ - _

.-

- _._

-.

.

.

-

.

i

O

.

nn

-

'<r<s9'

w .u

%.

. m ,s.,

, . . . , - _ . ,

..

-

1

.

1

i

OE Program

.

Assessment

e

l

Dick Andrews

e

44

.

, _ .

.

_

_

._

-

-

-._

.-- _ _ _

.

e

OE13rogram Assessment

.

i O

l

'

i..i

~

[

.I. . . '

.

'

'

, . . , .

.,

i o Focus on Use of Operating Experience.

.

'

101 Programs Evaluated.

!

To 3ics Reviewed.

I

- Issues warrant additional Management .

!

attention?

i

- Other locations known to be better?

.l0

- is industry experience being obtained and

j

utilized?

j

- is all other needed information being obtained?

l

- Are there barriers to using information obtained?

1

j

- Is information receipt adequate and being

4

properly utilized?

- Are personnel contacts and meeting participation

adequate?

- Have industry experts or peers provided input or

review?

- Is electronic information received adequate and

O

being utilized?

- Is printed information received adequate and

being utilized?

45

j

.

.

OE Program Assessment . .

e;

..... 2 " ="5 2 *L _

.

,

o Significant Results.

Oversight groups more aware of need to include

1

information utilization when performing audits,

surveillances, and assessment activities.

.

Heightened program owner awareness of ownership

and the need to use external as well as internal

g

information.

Discussion of nuclear operating experience at plant -

morning meetings has been implemented.

s Too many people rely solely on the formal OE

program for their external information.

.

Formal as well as informal industry sources of

information have been identified.

O

46

.

_

O

!

!

. J.ffy c . . ; . . . -- _ ; .

2.. f; , F,'

. ..

-

-

-

..

.<.,.m m m m y w .

. .n

..

-

.

) o 13 of 101 Programs Identifiec for

l

Additional Management Attention.

!

!

Protective Coatings

~

Setpoint Contro

-

Design Basis

!O

Radiological Consequences

l

Asbestos Management

l

ASME ISI/IST Program

j

Tagging

Chemical Contro

Hazardous Materials

Nondestructive Examination (NDE)

Bills of Materials

Offsite Dose Calculation, Radiological

O

Effluent Monitoring

M&TE

y

-

-.

-

-

. _ _ .

_

_

___

_.

O

OE Program Assessment

-

r -

o Programs identified for Technical

Review.

Erosion and Corrosion Programs

Fire Protection Program

PRA Program

Maintenance Rule Program

SG Inspection Program

g!

n ASME ISI/IST Program

AOV Program

MOV Program

Relief Valve Program

Check Valve Program

9 PM Program

Procurement Engineering Program

Fuel Reload Analysis

EEQ Program

SQUG Program

g

Control of Heavy Loads

Containment H Generation

2

48

- _

_

. _ _ _

. _ _ _ _ _ _

. .

j

!

l

!

.

l

i

o

,

4

l

h

' - ~i: l m;%&C%MGV , ifll<

.. m.a.u

... . . .

_ m , a m.w. . ..

I

!

I

.I

!

!

!

!

!

!,

l

Summar

and

.

,

t

l

OPPD Perspective

!o

l

)

'

Sudesh Gambhir

'

!

!

i

l

l

1

l

l

5

i

i

,! O

4

0

49

<

i

. - . .- _

i

-

Summary and

.

OPPD Perspective

y.n;3

.

o This Is a Serious Situation Which Must

l

Be Prevented and Not Repeated.

o Significant From an Industrial Safety

.

Standpoint.

o Significant From Plant Availability

Standpoint.

Not a nuclear safety significant event.

- Did not present a significant operational

challenge.

e!

50

_ _ -

-

_ _ _

_ _ _ _ - _

_

_

..

-

Maintenance Rule

.;O

Compliance

!

.

ver _

.

=

i

o Rule requirements were met
i

i

l

Piping was aroaerly monitored.

l

l

Appro ariate aerformance criteria were

j

established.

O

Industry wide operating experience was

utilized.

Monitoring was in place.

- With the exception of 4th stage extraction steam

piping, the program was effective.

o Further Assessment Is Planned to

Evaluate Implementation Against

" Excellence."

O

51

w

,

- _

_ _ _ - _ _ _ _ _ _ _ _

.

.

l

FAC Program

.

i

e

i

j

Jl2 *2L L.:

i

j o FAC program used for monitoring under

j

the Maintenance Rule.

l o Implementation of FAC program was

!

weak.

-

!

l o Industry experience was factored into

l

the inspecti.on program.

e

!

CHUG Database does not lead to

inspecting sweeps.

-

70% of sites in the extraction steam piping

l

were inspected prior to rupture.

o OPPD was deficient in utilizing are-1988

Fort Calhoun exaerience.

o FAC program limitations (Ref. EPRI doc.

I

NSAC 202, rev.1 page iv).

e

Tab 2

-

52

_

_ _ _ _.__

_ . _ _ _ _ _ _ _

-

.

l.

Assessments

.

i0

'

!

!

.

. w e a m w w o ,e ::

-

.

. .:. n .a

...;

.

~,m.~,-

.,s. . ,.

.

!

!

!

!

l

OPPD has conducted

~

!

multiple assessments to

0

!

ferret out the root cause

l

and any other program

i

!

deficiencies or problems.

!.

1

i

l

lO

l

53

.-.

_.

.

.

.

.

I

'

j

.

-

l

Corrective

-

l

Assessments

Actions & O

l

Results

.... _: C =

N"'

. . .

-

+ Lessons

Learned

. Updated

RCA

O

M

Short Term

i

Corrective

E+

o rec

lU4E

i C-

O

i

54

_

. . .

.

. . -

.

. -

. .

.

.

.

_ _ _

l

Other Considerations

.

O

!

!

l

.

..

m

,_ _

.

m v

s o a a ,:.ma,.

! o Corrective Actions

l

l

l

Extensive corrective actions including

i

consideration for generic impact.

.

i

,

o Historic Issue

lO

l

Problem occurrec because ofinadequate

treatment o" re alacement 3rior to 1989.

!

!

!

Missed op 3ortunity to inspect in 1985.

o Lessons Learnec

Lessons learned lave been shared with the

O

industry.

55

,

1

.

_

. _ _ _

_ _ _ _ _ _ _

- - - - - - -

l.

-

-

4

i

4

i

e

'

.

-l

}

-,

. ~ ~ 4' ?.' i,7MC . . ' '[ N '

-

.

.

. . , . . . . . . . .

. , . . . . . . . ,

i

!

1

>

,

l

Closing Remarks

~

,

i

i

e

i

I

s

a

Gary Gates

i

5

+

J

l

9

56

._

1

.

1

METHODOLOGY OF MONITORING EXTRACTION STEAM UNDER THE

MAINTENANCE RULE

^

White Paper in response to NRC Inspection Report 50-285/97-09

July 9,1997

1.

PURPOSE,

The purpose of this document is to explain the methodology used to determine

the scoping status of Extraction Steam components, the methods used to monitor

these components, and the basis used for determming the monitoring of these

components.

l

2.

REFERENCES

a)

NRC Inspection Report 50-285/97-09

b)

NEI 93-01 Revision 2, Industry Guidelines for Monitoring the

.

Effectiveness of Maintenance at Nuclear Power Plants

c)

NRC Regulatory Guide 1.160, Monitorine the Effectiveness.o_f

Maintenance at Nuclear Power Plants

d)

NRC Inspection Manual - Inspection Procedure 62706, Maintenance Rule

e)

NUREG-1526, Lessons Learned from Early Imolementation of the

Maintenance Rule at Nine Nuclear Power Plants

f)

Questions and Answers from the August 1993 NUMARC Maintenance

Workshops

g)

FCS Program Basis Document, Maintenance Rule

i

h)

FCS Maintenance Rule Implementing Instructions (MRII)

i)

FCS System Scoping Manuals (SSM)

'

j)

FCS PRA Summary Notebook

3.

MONITORING OF EXTRACTION STEAM PRIOR TO THE APRIL 21,

1997 EVENT

a)

Details of Extraction Steam Scoping

The Extraction Steam System (ESS) at Fort Calhoun Station (FCS) is

considered to be within the scope of the Maintenance Rule per

10CFR50.65(b)(2)(iii). These SSCs were included within the scope of the

rule since they can cause a plant trip. FCS had not experienced a plant

<

trip due to failure of extraction steam components, but industry experience

indicated that extrac+ ion steam could cause a plant trip at FCS.

Components within the ESS are monitored as part of the Feedwater

.

d

I

whitel.do::

--

--

..

'

.

.

Heaters functional groups.' The components within these functional

.

groups are not risk significant according to the plant PRA.2 The ESS is

not a Safety Related system, and there are no safety related functions for

extraction steam listed in the plant Design Basis Documents or the USAR.

Accordingly, the functional groups which include extraction steam are

classed as Non-Risk / Operating functional groups.

b)

Monitoring of Extraction Steam

I

i)

The functional groups containing extraction steam were monitored

using plant level criteria in accordance with 10CFR50.65(a)(2).

Guidance provided by NEI states that non-risk significant /

operating SSCs are to be monitored using plant level performance

criteria. This approach is endorsed by the NRC in Reg. Guide 1.160.'

'

ii)

Components within these functional groups have been monito. red

for failure by the plant NPRDS failure reporting process since

1991, when Revision 4 of the NPRDS Reporting Guidance Manual

(RGM) was implemented. The FCS Maintenance Rule Program

(MRP) adopted NPRDS failure reporting methodologies and

expanded NPRDS guidance to cover all components within the

PN

scope of the Maintenance Rule. Revision 5 of the NPRDS

Reporting Guidance Manual (December 1994) removed

requirements for monitoring many component types that require

monitoring under the Maintenance Rule. As a result, MRP

personnel created a Reporting Guidance Manual that supplemented

the NPRDS RGM and included instructions for continued

'

monitoring of through wall leakage of piping. These instructions

were later incorporated into Maintenance Rule Implementing

Instruction (MRII) -3, Maintenance Rule Failure Renortina. With

the cessation of NPRDS reporting, MRII-3 was revised to include

'

all necessary guidance for Maintenance Rule failure reporting, and

the NPRDS RGM is no longer used by the MRP as a reference.

iii)

In accordance with NEI guidance,5 a review of the failure history

of functional groups including extraction steam SSCs was

!

performed. This review consisted of two tiers. One was the

' Ref. i) Volume 15, Main Feedwater. Tab 10, LPA(B)HTR IPA (B)HTR HPFWHT - Feedwater

4

Heaters

2

,

Refj) 9.133.F.

' Ref. b) 93.2

d

i

Ref. c) 1.73

,

V

' Ref. b) 933

2

whitel. doc

'

.

.

review of existing plant specific NPRDS and Maintenance Rule

data from CHAMPS. The second tier involved the review of

Maintenance Work Documents from July 1,1992 through June 30,

1995 to determine if additional failure reporting other than that

already contained in CHAMPS was required. It should be noted

that NEI guidance only requires the licensee to perform a review

of history for a maximum of three years prior to the

implementation date of the rule.6 There were no problems

detected involving extraction steam that would require goal

setting.7 Based on these reviews, it was determined that the

functional groups containing extraction steam SSCs did not require

i

goal setting and were being correctly monitored in accordance

with 10CFR50.65(a)(2). While NEI guidance was used to make

this determination, the methodology used by FCS to determine the

proper monitoring level of extraction steam SSCs is dso discussed

in NUREG-1526 and endorsed in Reg. Guide 1.160.30

iv)

Extraction steam piping, as well as piping from other systems,"is

monitored by the plant's Erosion / Corrosion Program (ECP). The

ECP sets individual, component level performance criteria for

piping covered within its scope. The FCS MRP does not set

individual, component level performance criteria, but makes use of

existing programs as allowed by law" and NEI guidance. 2

^

Effectiveness of the ECP is monitored by the MRP using two

methods. These methods are discussed below:

a)

Failure Reporting Process - Through wall leakage of piping

is considered a component failure' per MRII-3. As such, a

failure investigation must be performed and a report

i

generated into the CHAMPS database when a leak occurs.

Since extraction steam SSCs are monitored at the plant

level, a non-catastrophic failure (leak) of an extraction

l

steam line would not exceed MRP performance criteria

unless plant level performance criteria (as described in

MRII-2, Settine Performance Criteria) were exceeded.

However, MRII-5, Component I ailure Analysis, allows for

the elevation of an SSC to monitoring under

.m

' Ref. b) 733

' Ref. i) Volume 15, Main Feedwater. Tab 15. Initial Performance Assessment

  • Ref. b) 9.2.4

' Ref. e) 2.4.1

Ref. c) 1.9

" Ref. c) B Use of Existing Licensee Programs

" Ref. b) 7.0

v

" Ref f) Appendix C, Section 12, Question 50

3

whitel. doc

- _ . _ _

.

_

_

_

._- .

. _ _ _

_ _

.

_

'

.

.

10CFR50.65(a)(1) even if performance criteria is not

.

^

exceeded, if the situation warrants.

b)

A catastrophic failure of extraction steam piping would

cause plant level perfonnance criteria to be exceeded (if

,

deemed an MPFF), or would exceed system level criteria

(Availability of 100% while power operation is desired)

even if the failure was not mairtenance preventable." In

effect, MRP performance criteria adequately monitors the

ECP since piping is monitored not only at the plant level

(as allowed by law), but also at the system and, to some

extent, the component level.

v)

A detailed analysis of the ECP by the MRP was not required as

part of the maintenance rule implementation effort. NEI 93-01

states:

Utilities can utilize their existingprogram results to

support the demonstration that SSCperformance is being

effectively controlled through preventive maintenance. Jf

verformance monitorine indicates that SSC verformance is

unacceptable. then the cause determination (Section 9.4.4)

e

verformed when SSC verformance is unacceptable should

correct any eauipment or vrogram deficiency. "

The FCS MRP was monitoring the effectiveness of the ECP, and

when the ECP was found to be ineffective in ensuring the

performance of extraction steam piping, a cause determination was

performed and goals were set as required by the rule.

c)

Conclusions

The following conclusions can be drawn about the status of the ESS prior

to the steam line break of April 21,1997:

i)

Extraction steam SSCs were correctly within the scope of the FCS

MRP.

ii)

Functional groups containing extraction steam SSCs were

correctly included for monitoring under 10CFR50.65(a)(2) based

on a review of maintenance history.

" Ref. i) Volume 15, Main Feedwater. Tab 2, Main Feedwater System

y

" Ref. b) 7.0

4

A v :. doc

_

_

'

.

.

iii)

A failure such as that occurring on April 21,1997 would have

-

^

caused both plant level and system level performance criteria to be

"

exceeded. MRIl-6, Placement of SSCs into Category (a)(1) or

(a)(2), would have prompted a cause determination as required by

law and NEI guidance. Such a cause determination would cause a

review of the effectiveness of existing programs as well as current

, performance criteria."If the ECP was fcand to be ineffective,

functional groups containing extraction steam SSCs would be

monitored under 10CFR50.65(a)(1) until effective corrective

l

action was taken. This meets the requirements of Reg. Guide 1.160" and NEI 93-01."

4.

MONITORING OF EXTRACTION STEAM AFTER THE APRIL 21,1997

EVENT

l

a)

Actions Taken Directly as a Result of the April 21,1997 Event

,

i)

The failure of a large radius sweep in the fourth stage extractio'n

line on April 21,1997 caused operations personnel to manually

trip the reactor. The failure of the pipe is classed as an equipment

failure per MRP failure reporting guidance." This requires an. '

investigation of the incident and the generatior, of a failure report.

'S

Failure reporting procedures require that a determination be made

.

as to whether the failure was maintenance preventable or not.

Industry experience is also to be used when making this

^

determination.20 The MRP reviewed the preliminary root cause

analysis (RCA) performed by plant personnel. The preliminary

root cause analysis determined that there were several

programmatic deficiencies in the ECP, and that the failure could

have been prevented. The failed section of piping was also sent

,

'

out for a failure analysis to two different vendors. Completion of

the MRP failure report is pending the results of these failure

analyses.

The failure of the piping caused a loss of the functions.1 groups

'

containing the failed pipe. The failure also caused a plant trip.

MRP personnel had intended to wait until the completion of the

failure analyses before performing a cause determination. This

would be necessary in order to determine if the failure was indeed

" Ref. h) MRII-6. Placement of SSCs into Cateeory (aYl) or (aV2),5.6

" Ref. c) 1.9

" Ref. b) 9.43 and 9.4.4

" Ref. b) MRIl-3djaintenance Rule Failure Reportine. 6.2

v

    • Ref. h) MRII-3. Maintenance Rule Failure Reportine.14.29

5

whitel. doc

..

.

. -

..

.

, .

an MPFF as stated in the preliminary RCA and therefore a

_

violation of plant level criteria. However, since system level

availability performance criteria were also exceeded (which are

applicable even if the failure is not an MPFF), a cause

2

determination was performed as required by NEI guidance and

Reg Guide 1.160.2 This cause determination concluded that the

-

ECP, in its present state, did not adequately monitor Flow

Accelerated Corrosion (FAC) suseptible piping per

10CFR50.65(a)(2). Therefore goal setting would be required

under 10CFR50.65(a)(1). Corrective actions already completed

were reviewed, and goals were set to insure completion of

additional corrective actions. The question of monitoring of the

ECP was asked of the NRC at an industry conference and the

following answer was given:

,

Programs are not in scope; only SSCs andperformance. . .

A firstfailure ofa SSC within the scope ofthe Maintenance

.

Rule due to errosion / corrosion would require an effective

-

cause determination and corrective action. A .second

l

failure ofthe same kind uvuld require goal setting and

monitoring.

.

^

In accordance with the above, the ECP was not placed into

category (a)(1); rather all FAC suseptible piping is currently being

monitored in accordance with 10CFR50.65(a)(1).24

25

i

ii)

In accordance with applicable guidance , the plant made use of

industry experience during the goal setting phase of the cause

'

l

determination. Personnel from other utilities were part of tha ECP

self assessment, and other industry experts were called upon to

assist with the failure analysis and to provide advice for program

'

improvements.

b)

Other Actions Taken as a Result of the April 21,1997 Event

.

A review was performed by MRP personnel to detennine if other piping

systems currently within the scope of the rule were being effectively

monitored by the ECP or by other preventive maintenance (PM)

2

  • ' Ref. b) 93.4

22 Ref. c) 1.7.1 Cause Determination

    • Ref. f) Appendix C, Section 16, Question 6

FCS MRP Cause Determination alfile\\cc249705 Revision 0

28

v

Ref. b) 9.4.1

6

whitel. doc

_

_

__

_

_

. . _ _ _

.__

\\

_

]

\\

-

programs.26 This evaluation resulted in the piping for three additional

-

"

^'

,

systems to be transferred from rnonitoring under 10CFR50.65(a)(2) to

!

I

monitoring under 10CFR50.65(a)(1). While these systems were being

monitored under system and functional group level performance criteria,

no evidence could be found that piping was being specifically monitored

.

'

in any way other than the MRP failure reporting process. It should be

noted. that monitoring of these systems was not ineffective since the

failure reporting process had already detected piping problems that had

"

contributed to unavailability that lead to the Instrument Air Compressors

being placed into category (a)(1) in January 1997.27

4

5.

FINAL CONCLUSION

.

Upon reviewing the documentation listed in Section 2 of this paper, and

j

reviewing the monitoring of extraction steam piping prior to the April 21,1997

event as well as corrective actions taken after the event, MRP personnel must

conclude that this event did not constitute a violation of 10CFR50.65. This .

conclusion is based on the following points:

a)

The SSCs involved were classed as non-risk / operating SSCs and

)

therefore monitoring at the plant level was appropriate.

A

b)

The Maintenance Rule only required a three year review of maintenance

history to determine the level of monitoring necessary for these SSCs. No

problems were detected during the review period. In addition, the

proactive replacement of several extraction piping components by the

,

ECP led MRP personnel to believe that the ECP was successful in

detecting piping deterioration prior to failure. As such, there were no

grounds for goal setting for these SSCs prior to the April 21,1997 event.

This is in accordance with 10CFR50.65(a)(2) which states " monitoring as

specified in paragraph (a)(1) ofthis section is not required where it has

been demonstrated that the performance or condition ofa structure,

system, or component is being efectively controlled through the

performance ofappropriate preventive maintenance, such that the

structure, system, or component remains capable ofperforming its

intendedfunction. "

c)

The purpose of the Maintenance Rule is to monitor the effectiveness of

preventive maintenance and take appropriate corrective action when

preventive maintenance is found to be ineffective. Paragraph (a)(1) of the

rule requires that "each holder ofan operating license . . . shall monitor

FCS MRP Analysis efar\\cc249706

FCS MRP Cause Determination alfile\\l4069701

v

7

whitel. doc

i

l

?

e

I

i

'

the performance or condition ofstructures, systems, or components,

~

against licensee-established goals, in a manner sufficient to provide

reasonable assurance that such structures, systems, and components .

are capable offulfilling their intendedfimetions. Such goals shall be

established commensurate with safety and, where practical, take into

account industry-wide operating experience. When the performance or

conclition ofa structure, system, or component does not meet established

goals, appropriate corrective action shall be taken. "

d)

Extraction steam piping at FCS was being monitored by the MRP at a

level that was in accordance with NEI 93-01, Revision 2, as endorsed by

NRC Reg. Guide 1.160. When these performance criteria were exceeded,

comprehensive corrective actions were taken and all requirements of NEI 93-01 were performed as to the re-catigorization of the SSCs involved.

e)

Industry operating experience was taken into account during the initial

scoping of extraction steam SSCs, and this experience resulted in the ,

inclusion of extraction stream within the scope of the Maintenance Rule.

Industry experience was also taken into account in the setting of goals

after plant level performance criteria had been exceeded.

f)

The failure in question is not a repeat MPFF. The failed component is a

large radius sweep, and there is no available industry information

m

indicating that these components are susceptible to accelerated FAC,

Preliminary failure analysis results also indicate that the flow rate within

the affected piping was considerably higher than expected due to the

configuration of the piping. NEI 93-01 defines MPFF and repeat MPFF as

follows:

An initial MPFF is thefirst occurrencefor a particular SSCf r

which thefailure results in a loss offunction that is attributa!!e to

a maintenance related cause. An ininial MPFFis afailure th:

would have been avoided by a maintenance activity that has n, t

been otherwise evaluated as an acceptable result (i.e., allowed sa

run tofailure due to an acceptable risk).

A " repetitive" MPFF is the subsequent loss offunction (as defined

above) that is attributable to the same maintenance related c.=?

that haspreviously occurred (e.g., an MOVfails to cine because a

spring pack was installed improperly -- the next tir.e this MOV

fails to close because the springpack is installed 'mproperly: the

MPFF is repetitive and the previous corrective c : tion did not

preclude recurrence). A second or subsequent loss offunction that

resultsfrom a different maintenance related cause is not

v

considered a repetitive MPFF (e.g., an MOVinitiallyfails to close

l

8

whitel. doc

i

.

.

1

.

because a springpack was installed improperly - the next time it

^

fails to close, itsfailure to close is because a set screw was

.

improperly installed: the MPFF is not repetitive). "

.

The current failure is the first functional failure of an extraction steam

pipe at FCS due to an ineffective ECP. It is also the first failure of a large

radius sweep due to FAC during the historical monitoring period required

by the rule. As such, it is an initial MPFF and not a repeat MPFF. Gther

i

extraction steam components (small radius elbows, reducers, etc.) were

being comprehensively monitored by the plant ECP, in part due to the

available industry operating experience considering these components to

be a problem. Since this is not a repetitive event, the failure in question is

not a violation of 10CFR50.65.

.

  1. f'%

-

.

L

g

Ref.b) Appendix B

9

whitel. doc

.

-

,

T;;tra Engineering Group, Inc.

t toucomeacsw sweet suite 800

r sephone (seos est 4622

e

Westogue, Conneco' cut 06069 USA

Fa phone (860) SS15524

,

r.

1

l

July 17,1997

97-FCA-203

Mr. Joseph K. Gasper

Omaha Public Power District

5 miles North of Fort Calhoun NE on Highway 75

Fort Calhoun Station

P.O. 399

,

Fort Calhoun, NE 68023 USA

Dear Joe:

Encimed please find the final version of Tetra Engineering Report TR-97-009 entitled " Fort

Calhoun Flow Accelerated Corrosion Assessment of Extraction Steam Line".

.

Sincerely,

Frederick C. Anderson

A

Vice President Engineering Services

e-mail: FAnderson_ Tetra @compuserve.com

-

'

v'

Engineering & Services forIndustry

. -

-

--

.-

. -

- - -

_ _ .

_ _ - - -

.

_

l*

-

-

I

-

TR-97-009

Fort Calhoun

Flow Accelerated Corrosion

Assessment of Extraction Steam Line

.

cT('g

Prepared For Omaha Public Power District

By Tetra Engineering Group, Inc.

July 17,1997

.

W Tetra Engineering Group, Inc.

USA: 110 Hopmeadow Straet, Suite 800, Westogue, CT, 06089 (1).860.651.4622

France:Immeuble Petra B, it.P. 272, 06905 SOPHIA AN11POUS (33).4.92.96.92.54

-

.

. _.

.

. . .

.

.

.

. . - - .

_

-.

-

. .

- -

W revo engineenne croup. Inc.

'

,7%

Contents

.

e

'

.

Introduction

1

,

BRT-Cicero Code

2

'

Fort Calhoun 4* Stage Extraction Steam Line

4

Comparison of BRT-CICERO to CHECWORKS

5

7

Conclusion

-

References

8

l

!

i

,

!

Appendix A

9

a,

--

I

l

<

f

1

l

l

.-

,

i

TR-97 009 Fort Calhoun FAc Assessment of Extraction Steam Line

Contents e i

.

L r,,,, ,,,,

, a,c

, c.

-

>

-

-

Introduction

On April 21,1997 a sweep elbow in the 4* stage extraction line of the Fort

Calhoun Nuclear Power Plant ruptured. The cause of the rupture was determined

to be flow accelerated corrosion.

A review of the flow accelerated corrosion program at Fort Calhoun was initiated

following the rupture. One facet of this review is a attempt to determine why the

existing program failed to identify the thinned component piior to rupture. The

Fort Calhoun program used the EPRI CHECWORKS computer code to identify

potential thinned components for inspection. An alternate methodology was

developed by Electricit6 de France for predicting flow accelerated corrosion in

power plant components. The EdF code is entitled BRT-CICERO.

This report contains the results of a BRT-CICERO analysis of the Fort Calhotm

4* stage extraction steam line.

een

-

l

.

l

TR-97 009 Fort Calhoun FAC Assessment of Extraction Steam Line

introduction .1

_

_

.

___

. _

_ _

'

Tstra Engine 1nng Group, Inc

4

-

-

.

BRT-Cicero Code

The BRT-CICERO' code was developed by M. Bouchacoun and F. N. Remy of

EdF/SEPTEN and J. de Toni of EdF/CNEPE. The intent of the development of

the code was to provide a centralized method for predicting and controlling flow

accelerated corrosion at the 54 nu: lear plants operated by EdF. The main

objectives of the code are the removal of the likelihood of a pipe rupture and the

reduction of maintenance program costs. The code is currently used by EdF to

l

diagnose the state of piping in the secondary system of a plant, assess component

lifetimes, prepare inspection campaigns, optimize replacement and repair

l

strategies, and document analysis,

The code provides a database function for the codes and standards used in the

'

plant construction, an inventory of the lines and elements, plant operating history,

plant chemistry history, and inspection results. Algorithms used to predict flow

accelerated corrosion of components are based principally on a modified form of

the Sanchez Caldera model a supported by testing" performed in the CIROCO

l

2

test loop run by EdF and feedback from operating plants. The basic predictive

~'

algorithm used in the BRT-CICERO code to determine the rate of FAC is as

follows:

i

'

FAC = f(Cr) * f(6) * (C,* - C.,)

(1)

,

g,

0.5 * ( k + D )i

<

Where:

f(Cr)

Alloy Composition Factor

f(0)

Oxide Porosity Factor

C,y

Equilibrium Soluble Ferrous Ion Concentration

C.,

Bulk Soluble Ferrous Ion Concentration

k

Mass Transfer Coefficient

6

-

Oxide Layer Diffusion Factor

D

,

The Alloy Composition Factor is a function of the Chromium, Molybdenum, and

Copper concentrations in the material ofinterest. BRT-CICERO assumes an

average alloy composition if no specific information is available. The average

values are based on extensive testing by EdF of a large number of heats of carbon

steel material. EdF also tests alloy composition of each component inspected as a

. . .

TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam Line

BRT-Cicero Code . 2

'

Tetra Engineenng Group, Inc.

i

~

matter of policy. This removes a considerable uncertainty in the FAC rate

<

1

.

predictions which would otherwise be present if the alloy composition is

unknown.

The Oxide Composition Factor and the Ferrous Ion Concentrations are functions

of pH and temperature. The Mass Transfer Coefficient 's a function of velocity

and the oxide Layer Diffusion Factor is a function of temperature.

Equation 1 applies to straight pipes. For elbows, tees, aad other geometric

discontinuities a geometry factor is applied. This geometry factor is a function of

,

the Sherwood number and accounts for the increased mass transfer as a result of

the discontinuity.

5

The BRT-CICERO code is used by first constructing a plant database. The plant

database consists of all susceptible lines modeled from isometric drawings plus

j

data such as construction code, design conditions, and operating conditions.

{

Additional information which must be entered includes pipe material properties,

nominal wall thickness plus tolerances, plant operating history, water chemistry

,

history, etc. Structural margins are then determined in order to quantify the

available wall for acceptable FAC degradation. Wear calculations are tlien

performed on all components and predicted wall thickness with associated

j

uncertainties determined for each component.

l

,

The code is used to assist in the selection of camponents for inspection.

l

Components can be classified in terms of the margin between projected thickness

and design thickness, the wear rate, or the time to minimum required wall

thickness. The code can then be used to determine the minimum inspection

frequency for the component.

When a component is inspected the 'UT information is entered into the code md

measured wear determined. Projections of the future wall thickness are based on

the observed wall thickness.

,

!

i

l

i

i

i

l

TR 97-009 Fort Calhoun FAC Assessment of Extraction Steam Line

BRT-Cicero Code . 3

!

__

_

. _

-__.. - .. _-__ -

.--

-

-

. -

. -

'

Tatra Enginernng Group. Inc.

-

.

.

Fort Calhoun 4th Stage Extraction Steam

Line

'

The BRT-CICERO code was used to assess the rate of FAC in the failed section

of the Fort Calhoun 4* stage extraction steam line using the data provide in

appendix A. Only the section of the line from the nozzle to the first tee was

considered. The following assumptions were made:

1. The Unit was assumed to be at 100% power for the entire 145000 operating

hours.

2. Chemistry was assumed constant at a pH of 9.44,

3. Fluid conditions in the line were:

'

Enthalpy 2.64x10' kjoule/kg

Flow

36.9 kg/see

Pressure

18.961 bar

4. Chromium content of the failed sweep elbow was 0.068%.

.

5. The radius of the sweep elbow was 1.5 m.

Based on the above input assumptions, the BRT-CICERO code projects a wall

loss of 10.9 mm or 0.429 inches for the 145000 operating period'. This compares

with the range of possible initial wall thickness of the component of 0.328 to

0.422 inches. The range ofinitial wall thickness reflects the nominal wall

thickness of the component *12.5% for the as procured tolerance.

The wear projected by the BRT-CICERO code is conservative to the observed

wall loss by anywhere from 0.057 to 0.151 inches depending on the actual initial

thickness of the component and assuming the rupture occurred with 0.050 inches

of wall remaining. This is reasonably good agreement given the assumptions

regarding operating conditions and chemistry.

1

)

TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineFort Calhoun 4th Stage Extraction Steam Line = 4

I

..

.

.. - -

-

-

_ _ _

..

-.

.- - . -

-.-

-

'

& Te*rs Engineenng Group. Inc.

'

.

.

!

Comparison of BRT-CICERO to

CHECWORKS

-

'

l

The CHECWORKS model of the Fort Calhoun 4* stage extraction line was not

available for review, therefore a detailed comparison of the CHECWORKS and

BRT-CICERO models could not be performed. However, some observations can

be made with regard to the general calculational approaches. For the purposes of

this report, only observations pertinent to the determination of the 4* stage

extraction line failure are provided.

There are three main differences in the approaches used by the two codes which

may have an effect on the prediction of wear in the 4* stage extraction line. These

differences are related to the use of a line correction factor, the treatment of alloy

content, and prediction uncertainties.

CHECWORKS employs an adjustment factor termed the "Line Correction

j

Factor"in the " pass 2" wear calculation. This correction factor is intended to

adjust the predicted wear rate of all components in a line by considering the

'

differences between the predicted wear and the measured wear for the components

,

on that line that were inspected. This has the effect of adjusting the predicted

'

wear of one component based on the measured wear of a different component,

essentially broadening the use oflimited inspection data. This is useful provided

that all components in the line are behaving in a similar fashion and the inspected

components contributing to the line correction factor are carefully selected.

The BRT-CICERO code does not employ a line correction factor. Inspection data

from one component is not :xtrapolated to other components. Inspection data is

used to adjust the predicted wear for that particular component only.

Both codes use a default value for alloy content of a component when no

information is available. CHECWORKS assumes a alloy content value of 0.0%

for carbon steel components, while BRT-CICERO uses an average value. The

0.0% value assumed by CHECWORKS would be conservative when calculating

the wear rate of an individual component. However, when coupled with the use

of a line correction factor, a potentially non-conservative scenario may occur. A

non-conservative prediction of wear could occur when a limited number of

examinations are performed on a line and the components selected happen to have

unmeasured alloy contents greater than the component ofinterest. The inspected

components would have lower wear rates due to the alloy effect. When the line

correction factor is used and wear for the line is determined, the projected wear

rate for the low alloy content component may be non-conservative.

.a

TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineComparison of BRT-CICERO to CHECWORKS . 5

-

bN ras engsneenne aroup. Inc.

'

.

Uncertainty in the prediction of wear is explicitly addressed on an individual

component basis in the BRT-CICERO code but not in the CHECWORKS code.

m

THE BRT-CICERO output provides an average wear value plus upper and lower .

bound values. The uncertainty in the predicted wall thickness is based on the

uncertainty in the initial wall thickness, which is typically 112.5%, plus

uncertainty in the alloy composition, uncertainty in UT measurements, and

uncertainty in the wear calculation. If a baseline inspection is performed or once

the comp' nent is inspected during service, the uncertainty in the initial thickness

o

is eliminated. Thickness projections and associated uncertainties are "re-zeroed"

j

from the inspected wall thickness measurement. This can not be done easily in

CHECWORKS for components that have no initial baseline and that are inspected

after an initial operating period. Instead, a nominal wall thickness is assumed and

wear is emnulative c,ver the life of the component.

Similarly, the contribution of the alloy uncertainty can be eliminated in the

BRT-CICERO code by the performance of an in-situ alloy analysis.

CHECWORKS also has the capability to record and apply the allow composition

j

should the composition of a component be determined. It does not, however,

'

explicitly address the uncertainty of unknown alloy compositions on a component

basis.

!%

.

-

..-

TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam LineComparison of BRT-CICERO to CHECWORKS e 6

. .

. _.~..

. . . . _ .

_ ._ - . _ . _ _ _

_ _ . _ . _ _

. _

..

. _ _ _ _

._ .

N ntre ene,neenne amuo. anc.

'

l

.

-

.

Conclusion

.

4

The BRT: CICERO code was used to model the failed section of the Fort Calhoun

.

!

4* stage extraction steam line. The predicted wall thinning from the code was

8

conservative to the observed degradation but still within reasonable agreement.

More important is that the predicted time to failure for this component would have

>

l

been less than the observed time to failure.

I

It should be noted that the exact measured chromium content was used in the

determination of the wear for this component. This in ormation was not available

r

prior to the rupture. The value of 0.068% is somewhat less than an average of

m

approximately 0.16W and may have contributed to the high wear rate. If an

l

average value for chromium was used in the BRT-CICERO code a somewhat less

conservative wall loss would be predicted. This analysis was not done, but it is

likely that the code would have projected a failure of this component in time to

1

l

avoid the actual rupture.

l

i

,

}

-

a

k

s

J

TR-97-009 Foot Calhoun FAC Assessment of Extraction Steam Line

Conclusion * 7

. . . . _ _ .

.

.

_.

.-.

_ . _ . _ _

.

. _ .

.

. . _

-

W retis engineenne roup, Inc.

'

c

.

.

.

.

,

e

References

I

]'

l'.

Bouchacourt, M.,"BRT-CICERO 'A Software for Controlling Flow

Accelerated Corrosion' - User Manual," Revision A, EdF, Lyon, France, June

i

1995.

2. Sanchez Caldera, L. E., "The Mechanism of Corrosion-Erosion in Steam

j.

Extraction Lines of Power Stations," Ph.D. Thesis, Massachusetts Institute of

Technology,1984.

3. Cragnolino. G., Czajkowski, C., Shack, W. J., " Review of Erosion-Corrosion

j

j

in Single Phase Flows," NUREG/CR-5156, April 1988

i

4. Ducreux, J., "The Influence of Flow Velocity on the Corrosion-Erosion of

,

Carbon Steel in Pressurized Water," Water Chemistry 3, BNES, Lond'on,

l

1983.

5. Berge, P, Khan, F.," Corrosion-Erosion Des Aciers Dans L' Eau et la Vapeur

'

Humide," R6 sums et conclusion de la reunion de sp6cialistes, Mai 1982.

i

6. M. Bouchacourt E-Mail to F. Anderson," Transmittal of BRT-CICERO code

,rs

results", June 2,1997.

]

I

i

7. Jonas, O., " Erosion-Corrosion of PWR Feedwater Piping Survey of

Experience, Design, Water Chemistry, and Materials," NUREG/CR-5149,

March 1988.

4

1

1

,J

TR-97-009 Fort Calhoun FAC Assessment of Extraction Steam Line

References . 8

..

.

. _ _ _

. _ .

._

-

.

.

i

s

.

,

i

e**

OEP7blicP7,verD$

~

"

444 South 16th Street Mall

Omaha NE 68102-2247

June 4. 1997

LIC-97-0087

i

U.S. Nuclear Regulatory Commission

Attn: Document Control Desk

'

Mail Station P1-137

Washington. D.C.

20555-0001

References: 1.

Docket No. 50-285

2.

LER-97-003 Manual Reactor Trip Due to a Steam Line. Rupture

SUBJECT:

Assessments Related to the Extraction Steam Line Rupture of April

21, 1997

l

'

As committed in the May 5. 1997 Public Meeting, please find attached the

assessments completed in response to the extraction steam line rupture of

)

April 21. 1997.

These documents contain Omaha Public Power District's (OPPD)

internal findings and recommendations concerning the Extraction Steam Line

,

Rupture event that occurred at Fort Calhoun Station (FCS) on April 21, 1997.

OPPD's corrective actions for this event are listed in LER-97-003.

For the

purpose of providing additional detail to the NRC the corrective actions in

LER-97-003 are expanded upon in Attachment 1 of this correspondence.

However.

'

these specific actions may change as OPPD continues to review and improve its

program and are not meant as additional commitments.

At this time, the failure analysis of the ruptured elbow has not been

received. This item will be sent at a later date.

Please contact me if you have any questions.

Sincerely,

y&

S. K. Gambhir

Division Manager

Engineering and Operations Support

v

c5.5124

Employment with Equal Opportunity

-

_ _ _

_._

_ _ _ _ _

- _ . _ . _ _ . .

_

_

_. .

._

, .

_

!

l.

s

Attachments 1.

Additional Information on Commitments to the NRC for

Co~rrective Actions Listed in LER-97-003

,

2.

Fort Calhoun Station Root Cause and Generic Implications

'

Report Fourth Stage Steam Extraction Line Rupture CR

199700445 Revision 0

,

3.

Damage Assessment Report for the Break in the Extraction

j

Steam Line Revision 0. May 3, 1997

,

4.

Fort Calhoun Station Erosion / Corrosion Program Assessment

'

Report, dated May 2. 1997

i

5.

Fort Calhoun Station Self Assessment Erosion / Corrosion

Program Team Findings, dated May 6. 1997

SKG/ddd

.

j

c:

Winston & Strawn

E. W. Merschoff, NRC Regional Administrator. Region IV

'

'

'

J. L. Shackelford. Senior Reactor Analyst. DRS

l

L. R. Wharton. NRC Project Manager

.

W. C. Walker. NRC Senior Resident Inspector

1

.

.

"

g

$

"

>

1

a

)

,

.

~.)

, _ - .

_ ._ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ ._. _ ._ _ . _ _ __ _ ..- _ __. _ . ____ - _._ -__ ..__ ____

j

-

.

!

i

Attachment 1

'

-

i

-

l

Additonal information on Corrective Actions

!

Listed in LER-97-003

$

i

4

-

d

'!

t

i

1

'

r

$

l

4

1*

.

4

.

J

r

i

.I

}

i

i,

4

,

!

"

j

..

.

!

!

1

i

-

.

i

.

.

!

\\

l

1

~

i

i

1

!

!

i

.

-

/

4

4

4

.

6

nv-

e,

,

v

-

- - - - - - -

- - - - ,

..-

-,e-

,

, -.

-e,,-

. _ . . _ . _ _ _ _ _ _ . _ _ - _ . _ _ _ _ _ _ . _ . _ . . _ _ _ . _ _ _ . _

j

-

-

.

!

d

l

Attachment 1

'

-

-

b

Additional Information on Corrective Actions Listed in LER-97-003

)

!

In LER-97-03 OPPD committed to " Revise the Erosion / Corrosion Program Plan.

controlling procedures and modules to be consistent with industry standards.

!

This revision will include upgrade of the implementing procedures to be

consistent with industry standards (e.g. NSAC 202L. Rev.1), development of

susceptibility documentation and requirements for use of current industry

i

experience. This will be completed by the beginning of the 1998 Refueling

i

j

Outage." This commitment will address the following issues identified in the

self assessment:

l

1.

Procedures should include more specific guidance on how Outage

i

inspection locations ~are chosen.

i

1

i

2.

Measured wear determination process should incorporate the following

industry practices:

.

1

-

'

j

a.

Use of accepted practices to determine lifetime component wear

j

(circumferential band, moving blanket, point to point).

i

l

b.

Clarify the use of engineering judgment relative to wear

determinations.

j

-

,

.

c.

Process for incorporating measured wear into CHECWORKS models.

-

l

.

.

,

3.

Sample expansion process should be revised to align it with industry

i

standards.

Specific changes include:

l

-

i

a.

Clarify wording for small bore piping.

l

b.

Add requirement for inspecting upstream of expanders / expanding

elbows.

2

c.

Clarify that pressure / temperature exemption only applies to raw

water systems.

d.

Clarify that expansion is to parallel components in each train.

~

e.

Define the terms " component" and " highest wearing".

7

Specify highest wearing components in the same train.

4.

A document is needed to describe and control the identification of

susceptible systems.

.

1

-

-

-- .

._

.

_.

-

. - - - -

..

.

_

_ -

-.

.

.

5.

A document is needed to describe and control the evaluation of

~

'

- . .

susceptible systems.

,

6.

Documentation should be provided when grid refinement.or scanning is

3

performed when significant erosion / corrosion is found.

i

7.

The inspection data evaluation for components with PASS 2 CHECWORKS

.

analysis should consider current predicted wear rates.

'

l1

8.

The program should employ verification of many elements and a formal

process for changing program elements needs to be established.

(Examples: CHECWORKS model changes. Erosion / Corrosion Program General

Information Table. Erosion / Corrosion Program Technical Data Review.)

i

.

Also in LER-97-03 OPPD committed to " Revise and verify the Fort Calhoun

CHECWORKS models consistent with industry standards by December 31. 1997."

i

This commitment will address the following issues identified in the self

assessment:

4

.

.

1.

Current plant CHECWORKS models need to be verified.

2.

Inspection data from 1995 and 1996 outages needs to be incorporated into

j

CHECWORKS models consistent with industry practice.

A

3.

The plant CHECWORKS models need to be updated and controls put in place

to document changes.

-

In addit 1Bn OPPD plans to conduct a follow-up self assessment of the

erosion / corrosion program following the next refueling outage to evaluate the

.

effectiveness of program enhancements.

.

$

During the implementation of our corrective actions OPPD will review and

j

incorporate, as appropriate, the following recommendations of the self

assessment team:

1.

Program Basis Document should be updated to:

a)

Eliminate duplication.

b)

Remove unnecessary detail.

c)

Make Program Basis Document and inspection procedure consistent.

2.

Program Basis Document should describe how susceptible systems are

dispositioned with respect to analysis.

.

-s

2

_

.

. . .- . ._.

.--

-

- - - - - - -

-

--

- - - - - - - -

.

.

l

-

3.

Program Basis Document should be updated to enhance communication

'

responsibilities between Operations and Maintenance regarding feedback

j

to Erosion / Corrosion Engineer.

4.

Guidance for documenting rationale for inspecting a specific location

should be provided.

5.

The inspection data form should be enhanced to address the following:

a)

Directions for the use of this form.

b)

Direction for what to do when measured minimum wall thickness is

inconsistent with nominal wall thickness.

c)

Direction on what to do when previous thickness is not available

and nominal wall thickness is not used.

d)

Clarify the step " List expansion of test sites not previously

evaluated."

-

6.

Typographical errors in the Program Basis Document should be corrected.

7.

Typographical errors and inconsistencies in the inspection procedure

should be corrected.

m

8.

The highest allowable value of maximum allowable stress (SE) should be

used to eliminate unnecessary conservatism and ensure consistency with

CHECWORKS.

9.

Program documents should be revised to proceduralize the following:

.

a)

Trending of inspection results.

b)

Qualitative evaluation of inspection data.

c)

Evaluation of data to ensure thinning is bounded.

10.

The acceptance criteria (Design Minimum Wall) being used for inspection

data evaluations should be reviewed to ensure that OPPD applicable code

requirements are met.

.

11.

The grid size on 6" components should be reduced to comply with

recommendations in NSAC 202L.

12.

A clear separation in Erosion / Corrosion Program documentation between

Flow Accelerated Corrosion (FAC) and other wall degradation mechanisms

(such as raw water corrosion) should be provided.

.

3

-

-

-

.

.. - - .

- _ -

.

-

-

.

- . .

. -..

'

.

,

i

.

.

13.

The guidance of NSAC 202L should be employed to address susceptible

,

piping that is not suitable for modeling.

14.

Complete CHECWORKS models of susceptible piping that is suitable for

modeling should be developed.

15.

The isometric-drawings identifying CHECWORKS component identifiers are

vital tools.

A set of these drawings should be placed in retrievable

storage and a second set should be used as a working copy.

16.

Sample expansions should not rely excessively on older inspections and

.

should aggressively seek to ensure thinning is bounded.

17.

Sample expansion on tee branches need to be performed on the train

'

,

containing the branch.

18.

The inspection data evaluation process should address how to handle

,

components with readings greater than nominal wall thickness.

-

'

-

19.

Any exceptions to the grid procedures should be noted on the layout

diagram provided in the outage summary notebook.

20.

The CHECWORKS Program should be used to perform inspection data

m

evaluation.

21.

Each inspection data package should include a printout showing the

inspection data matrix.

22.

Although it appears that informal communication does exist between

.

various departments and the Erosion / Corrosion Engineer the

Erosion / Corrosion Engineer should perform a review of emergent KdRs via

the Daily Emergent list.

Review of this list should give a heads up to

any developing system abnormalities.

23.

The closure review of configuration change and maintenance documents

should be strengthened to identify any issue of concern to the

Erosion / Corrosion Engineer.

'

l

24.

The Outage Scope Change / Addition Request form should be revised to

ensure that requests for deletions and additions to outages are properly

evaluated for Erosion /Cerrosion scope.

l

l

25.

Feedwater iron transport information should be added to the Seconday

Chemistry Monthly Summary Report.

26.

OPPD is a member of the CHECWORKS Users Group (CHUG). but is not an

4

1

.

i

d

active participant.

FAC personnel, especially the Erosion / Corrosion

'

Engineer, should attend the CHUG meetings. These meetings are held

twice a year and cover current Erosion / Corrosion technical issues as

well as a forum for discussing plant experiences.

27.

OPPD should participate in the CHUG Plant Experience Database to ensure

Erosion / Corrosion staff obtains future updates to this important

industry experience database.

_

28.

As a one nuclear unit utility, it is recommended that OPPD consider

joining with a group of other similar plants at other utilities to form

a peer group to share experience and peer assistance.

I

29.

A program to provide flow accelerated corrosion sensitivity training to

l

applicable plant staff beyond Erosion / Corrosion personnel should be

considered.

This will help to ensure that plant conditions that may

affect flow accelerated corrosion are communicated to the

Erosion / Corrosion Engineer and incorporated into the Program.

4

. .

30.

Use of resistant materials and systematic replacements should be

i

considered.

[

31.

The program should incorporate management involvement in important

program elements.

(Examples: CHECWORKS models, outage inspection scope.

!

Rg

outage close-out)

'

'

,

_

.

.

g

~~

5

l

_

. _ _ . _ _ . _ . _ . _ _ . . _ _ . _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . - - -. _ _._____._

i

,

.

!

!

l

Attachment 2

'

,

l

Fort Calhoun Station Root Cause

l

and Generic Implications Report

4

Fourth Stage Steam Extraction Line Rupture

i

CR 199700445 Revision 0

l

l

,!

!

!

l

-

1

i

!

it:.

'

-.

.

'

,

-

.

!

,

!

9

,

4

l

!

!

1

4

4

1

4

b

!

<

i

i

Dfj3

y.

-- --

--

-e

-

ee--

.

. . _ .

_

.

__

_

. _ _ . _ _ . _ _

_

_ . . _ _ _ . .

. _ _ _ _ _ . _ _

.

1

1

_

-

FORT CALHOUN STATION

,

.

ROOT CAUSE AND GENERIC IMPLICATIONS REPORT

FOURTH STAGE STEAM EXTRACTION LINE RUPTURE

CR 199700445

PRC RECOMMENDS

REVISION 0

APPROVAL'

i

SRG-97-026

MY 0 71997

,

'

t>HG MTG.-MINUTpgg

/

A. R. Patel, Lead Evaluator

,

Date

ShT7

,

[. R. Geschw nder, Evaluator

'

Date

-

W,

4~Y7

fK. G' asper, P/er Review Team Member - CR Owner

~

V ' ate

D '

"

'

4]

e

1

G~/5/97

. L. Skiles, Peer Review Team Member

bate

..

.

R 4. M

    • /s*/r'

R. L. JaIvorski, Peer Review Team Member

.

Date

/

f d'99

R. L. Phelps, Peer' Review Team Member

Date

b

Sf0Cf97

R. L. Andrews, Peer Review Team Member

' Date

%)&

s/sh7

M. T. Sweigart, PeeVReview Team Member - NSRG

D$te

dukb b

skk7

e su,

M. Kellams, supeNisor - HPES/RCA

Ddte

'

- . - - - . _ . - . _ _ - . - - - - - - _ _ . - . - . _ - - - _ _ . - . - . - _ . - . . . - . _ - . -

-

.

.

l

Attachment 3

'

!

-

!

Damage Assessme::t Report.

for the Break in the Eyiraction Steam Line

i

j

Revision 0, n'4ay 3,1997

-

.

l

i

!

!

!

!

.

I

!

!

!

4

!

I

i

l

i

{h

'%

i

!

..

.

.

1

i

!

i

!

!

!

i

-

b

.

.

1

i

1

$

i

!

!

!

i

i

f'

..

.

.

. - .

.

.

.

_._

-

.

.

.

DAMAGE ASSESSMENT REPORT

'

-

.

-

.

FOR THE BREAKIN THE

.

)

EXTRACTION STEAM LINE

..

Revision 0

l

.3

-

_.

May 3,1997

.

_

i

R. L. Phelps, P.E.

i

Manager - Station Engineering

M. R. Core, P.E.

-

,

,. Manager- System Engineering

- _ _ - _ _ _ - - - - _ - _ _ . _ _ - - . - - - _ . - - . -

_ - _ . - - . - - - - _ _ .

- - . -

i

.

.

.

i

!

.

!

-.

Attachment 4

'

4

i

!

!

Fort Calhoun Station

I

Erosion / Corrosion Program Assessment Report,

i

i

-

dated May 2,1997

-

l

4

!

i

!

l

i

i

1

3

i

}

l

l

i

A

..

.

I

f

,

!

!

.

l

1

/

i

i

_

-

l

<

4

i

!

i

l

1

I

4

1

1

?

-

!

k

W "b >

' D

. .

.

.

.

.

.

.

.

fp,

.

_ - - _ _ - _ _ , _ . . _ _

-

.

.

FORT CALHOUN STATION

EROSION / CORROSION PROGRAM

-

'

ASSESSMENT REPORT

.

.

.

_

.

l

m

-

Mt

> 7

L

P. Hopkins

b' @~

Craig Y%n[r (%P)

- .

C OS&

Ned Dietrich (DE&S)

e

Gus Undall"(Representinh EPRI)

,

CSM 4r Dw cl A. A.4k

.

David Smith '(DPC):

Approved:

%<1A

Me K.'GasperCo-TeamLeader/ 'Taylof

1

. o-Team Leader

.

tek

d

Q

i

ack L. Skiles

Sudesh K. Garrbhir

Co-Team Leader

  • Sponsor

!

.

.

I

Attachment 5

'.

-

i

-

.

.

i

Fort Calhoun Station Self Assessment

j

Erosion / Corrosion Program Team Findings,

-

_

I

t

4

4

h

i

t

!

,

!

.

9

l

%.

I

-s

l

l

-.

.

!

!

-

,

l.

!

i

I

1

i

4

l

-

I

i'

i

i

i

!

,

t'

-

i

t

l

!

?

-

.

. . . -

.-

.- .

.

-

-

-

.. .

.

.

.

(~

j

R6 May L.

1997

Fort Calhoun Station SELF ASSESSMENT

Erosion / Corrosion Program Team Findings

Start Up Issues

!

Responsible

Corrective Actions

Completion:

Group

Date

1. Start Up Issues

A.

Upgrade the susceptibility evaluation to ensure

B. Lisowyj

A Restart FAC Susceptibility

05/08/97

(Should be

the following:

R. Aleksick

Review was performed to define

Complete

addressed prior

a)

Susceptible systems are covered.

D. Rollins

the scope of the FAC

Closure

to restart from

b)

Susceptible segments of susceptible systems

susceptible piping. As a

Memo

current Forced

are identified.

result the following

FC-0019-17

Outage.)

c)

Operators input is incorporated.

additional lines were added to

Section 2.0, F3

d)

Current susceptibility criteria are applied.

the program: Seal Steam

e)

Susceptible segments of susceptible systems

(entire system). S/G Blowdown

need to be addressed.

(suction / discharge of BD

transfer pumps). Condensate

Provide NRC the analysis and justification for

Recirculation (recirc.). Steam

not test.

Traps and Drains. Complete

any additional needed

inspections Pre Start-up.

Continue review to include

lines such as small bore.

Section 4.0, F1

B.

Review systems to ensure piping and components

R. Ruhge

Review MW0s. Mods. past

05/05/97

downstream of replaced components have been

R. Frakes

inspection data to determine

Complete

inspected to ensure industry experience has been

K. Hyde

inspected locat, ion. Perform

Closure

addressed. Document report.

an inspection not previously

Memo EOS-

Followup justification of why exclude in the past

completed. MWO 971649

SSE-97-065

to provide to NRC later.

(Complete)

Section 4.0. F2

C.

Component S-56 appears to have been installed

R. Jaworski

Inspect 5-56 and documentation

05/08/97

without the required reinforcing pad.

K. Woods

to determine thickness of

Complete

(Documentation is being pursued by OPPD personnel

component. Modify if

Closure

.

that may resolve this issue.) (Prior to Critical)

necessary.

Memo EOS-

ECN 97-161. CWO 97-037 (Comp..)

SSE-97-066

'

.

.

.

.

e

.

.)

R6 May '

1997

Fort Calhoun StoCion SELF ASSESSMENT

Erosion / Corrosion Program Team Findings

Start Up Issues

Responsible

Corrective Actions

Completion

,

'

Group

Date

Section 4.0. F3

D.

Packages from the 1996 Outage have'not been

D. Rollins

Review pkg 5-4. S-33. 0-26A.

05/08/97

l

independently reviewed (Examples: S-4. S-33.

A. Patel

S-38. D-84A. D-213 and other

Complete

D-26A. S-38. D-84A. D-213) as of 4/30/97. (Prior

pkgs if identified.

Closure

i

to Critical)

Memo

,

FC-0020-97 i

Section 4.0 F4

E.

Components displaying significant wear should be

N. Dietrich

Re-evaluate-components using

05/09/97

re-evaluated using industry standard techniques.

industry techniques. Identify

Complete

(Examples: S-73. S-74. S-66 and S-63)

needed inspections as needed.

Closure

Provide Technical discussion.

Memo

FC-0022-97

Sectior 4.0. F7

F.

Data or evaluations could not be found for the

D. Rollins

Locate documentation and

05/09/97

some 1996 inspection locations. (Examples: 5-57

R. Ruhge

include in database.

Complete

S-80. 5-92) (Prior to Critical)

Closure

Memo EOS-

SSE-97-068

+

i

i

l

'

.

.

.

.

. _ . _ . _ - _ . _ . . . _ . _ _ _ . . . . . _ . _ . . _ _ _ . _ _ _ _ _ _ _ _ . _ _ .

_

.

.i)

R6 May

'1997

b'

FortCalhounStationSELFASSESSMENT

Erosion / Corrosion Program Team Findings

!

Start Up Issues

,

'

Responsible

Corrective Actions

CompletionI

,

Group

Date'

!

Section 4.0. F9

G.

A review of high priority systems (Feedwater.

D. Rollins

Review the 6 hish priority

05/08/97

L

,

Steam Dump and Bypass. Blowdown. Extraction

B. Lisowyj

systems and identify points

Closure

Stean Condensate, and lleater Drains) should be

needing inspection. Complete

Memo

!

,

performed to ensure locations that industry

inspections or verify that

FC-0021-97 '

experience has shown to be potentially.

inspections have been

t

susceptible have been addressed.

performed. Complete necessary

Meeting on how we came up with these system and

repairs / replacements as

selection criteria.

necessary.

Inspection MWO's:

!

Complete:

i

971627(ES-3A). 971629(ES-2E).

I

971630(ES-2C). 971632(HD-3A.

>

38. 3C). 971649(11D-18).

971674(HD-IH) 971703(11D)

971666(SGB-2C)

Repair NWO's:

Complete:

I

971650(HD-3A). 971651(11D-38).

!

971652(llD-3C). 971655

.

t

i

!

B

'

t

!

.

t

.

.

.

s

___

._ _ - _ _ - - - - _ _

- -

_ _

m

..

_

._

_.

. _

.

.

i

FsrtCr,thouni.a$

1

R8 Hi

997

+

,

ion SELF ASSESSMENT

Erosion / Corrosion Program Team Findings

short Term / tong Term issues

Responsible

Corrective Actions

CceptetIJn

Group

Date

2.

Short 1erm issues

A.

Procedures should include more specific guidance on how

B. Lisowyj

Provide additional specific guidance

1998

(Should be

Outage inspection locations are chosen.

I

Procedure

for determining outage inspection

Refueling

addressed prior to

Group

locations in the Program Basis

Outage e

restart from the

Docunent (PD8) and other program

,

1998 Refueling

documents as needed.

Outage.)

Srction 1.0, F1

CID 970567/01

Szction 1.0, F2

s.

Measured wear determination process should incorporate the

8. Lisowyl

Incorporate industry practices for

1998

following industry practices:

Procedure

measured wear determinations in

Refueling

a)

(Jse of accepted practices to determine lifetime

Group

appropriate program documents as

Outage

comonent wear (circunferential band, moving

follows:

blanket, point to point)

a)

use of accepted practices to

b)

Clarify the use of engineering judgment relative to

determine lifetime component

wear determinations.

wear (circumferentist band,

c)

Process for incorporating measured wear into

moving blanket, point to point)

CHECWORKS models.

b)

Clarify the use of engineering

judgment relative to wear

determinations.

,

c)

Process for incorporating

measured wear into CHECWORKS

models.

CID 970567/02

Section 1.0, F3

C.

Sagte expansion process should be revised to align it with

8. Lisowyj

Revise the sagte expansion process to

1998

Industry standards. Specific changes include:

be consistent with industry standards

Refueling

a)

Clarify wording for smatt bore piping.

as follows:

Outage

b)

Add requirement for inspecting upstream of

a)

Clarify wording for small bore

expanders / expanding elbows,

piping,

c)

Clarify that pressure /teverature exemption only

b)

Add requirement for inspecting

applies to raw water systems.

d)

Clarify that expansion is to patattet co m onents in

upstream of expanders / expanding

elbows.

'

each train.

c)

Clarify that '

e)

Define the terms " component" and " highest wearing."

f)

Specify highest wearing co monents in the same

pressure / temperature exemtion

only applies to raw water

train.

systems,

d)

Clarify that expansion is to

parattel components in each

train.

e)

Define the terms "co monent" and

,

" highest wearing."

.

f)

Specify highest wearing

components in the same train.

CID 970567/03

.

-

. -

.

.

.

.

.

. .

.

.

.

.

.

. .

.

.

.

.

.

.

__

m

. _ _

.

.

t

R8 Mr

s e997

i

fort Csthounl

T,lon SELF ASSESSMENT

,

'

Erosion /Corrostw/>rogram Tsam Findings

Short Term /Long Term Issues

Responsible

Corrective Actions

Completion

Group

Date

-

S2ction ?.0, F1

D.

A docunent is needed to describe and control the

S. Lisowyj

Revise existing documents or deveicp

1998

Identification of susceptible systems.

'

Group

the process for the identification of

Outage

Procedure

new docments to describe and control

Refueling

,

'

susceptible systems.

,

CID 970567/04

,

Section 2.0, F2

E.

A docunent is needed to describe and control the evaluation

B. Lisowyj

Devel>ip or revise existing procedures

1998

.

of susceptible systems.

Procedure

to describe and control the evaluation

Refueling

Group

lofsusceptiblesystems.

Cutage

CID 970567/05

Section 3.0, F1

F.

Current plant CHECWORKS models need to be verified.

9. Lisowyj

Verify the current CHECWORKS modet.

1998

Refucting

CID 970567/06

Outage

Section 3.0* F2

G.

Inspection data from 1995 and 1996 outages needs to be

B. Lisowyj

Input inspection data from 1995 and

Prior to

Incorporated into CHECWORKS models consistent with industry

1996 outages into CHECWORKS n'.odel.

1998

practice.

Refueting

CID 970567/07

Outage

Stetton 3.0* F3

H.

The plant CHECWORKS models need to be updated and controls

B. Lisowyj

Ltpdate plant CHECWORKS model and

Prior to

1

put in place to doc m ent changes.

Procedure

revise existing documents or develop

1998

Croup

new docments to adninistratively

Refueling

contret revisions to CHECWORKS model.

Outage

Clu 970567/08

Srction 4.0* F5

3.

Docunentation should be provided when grid refinement or

B. Lisowyj

Revise existing docunents or create

1998

scanning is performed when significant erosion / corrosion is

Procedure

new documents to require documentation

Refueling

'

found.

Group

when grid refinement or scanning is

Outage

performed when significant

erosion / corrosion I,s found.

CID 970567/09

Section 4.0' F6

J.

The inepection data evaluation for components with PASS 2

B. Liscwyj

Revise inspection data evaluation for

1998

CHECWORKS analysis did not consider current predicted wear

components with PASS 2 CHECWORKS

Refueling

rates.

analysis to consider current predicted

Outage

wear rates.

.

CID 970567/10

.

.

.

__ - __-

.

. - - - - _ _ - _ - - _ - - - _ _ _ - - _ _ - _ - _ - _ _ _ - _ - _ _ - - - _ _ _ _ _ _ _ - _ - - - - _ _ - - - _ _ _ .

_ - _ - _ _ - _ _ _ - - _ . _ _ _ _ _ _ _ - _ _ _ . - - - _ _ - _ _ . _ - - _ . . _ _ _ . _ - - . - _ _ _ _ _ _ _ , - _ - -__

.

.

.

/'

,

E)

RS >.

I, i997

Fe:rt C:lhoun ..ttion SELF ASSESSMENT

Erosicn/ Corrosion Program Team Findings

Short Term /Long Term Issues

Responsible

Corrective Actions

Completioni

Croup

Onte

Section 4.0, F8

K.

The program does not appear to employ verification of many

8. Lisowyj

Revise doctanents or create new

1998

elements and a formal process for changing program elements

Procedure

documents to strmgthen adsinistrative

Refueling

does not appear to exist. (Examples: CHECWORKS model

Group

control of verifir.ation and changes to

Outage i

changes, Erosion / Corrosion Program General Information Table,

program elements (Examples:

Erosion / Corrosion Program Technical Data Revleu.)

CHECWORKS model changes,

Erosion / Corrosion Program Generat

Information Table, Erosion / Corrosion

Program Technical Data Review.)

CID 970567/11

3.

Long Term tstues

A.

A follow-up assessment should be performed following the next

J. Casper

Perform a format assessment after 1998

Post 1998

S ction 9.0, F1

refueling outage to evaluate the ef fectiveness of program

RF0 to evaluate the effectiveness of

Refueling

enhancements

program enhancements

Outage

CID 970567/12

I

.a

e

4

4

_ _ _ _ _ _ _ _ _ . . . _ . , _ _ _ _ _ . _ . _ _ _ _ _ . . . _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

-

. _.

.

. - -

. - _ _

_

- _.

.

.

l

Enclosure 3

-

Fort Calhoun Station

NRC Handout

i

,

t

i

.

4

, . . . . - . - . . - - . . - -

. . . . - - . . ~ . . _ .

. - . - . - - . . - . - . - - ~ . _ . - .

_ - . _ . - - . . .

).

a

. . .

.:

..

,

PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA

4

.

'

CONFERENCE WITH OMAHA PUBLIC POWER DISTRICT

,

July 21,1997

.

i

NRC REGION IV, ARLINGTON, TEXAS

,

4

1.

' INTRODUCTIONS / OPENING REMARKS - Ellis Merschoff, Regional Administrator

2.

ENFORCEMENT PROCESS Michael Vasquez, Enforcement Specialist

'

i'

3.

APPARENT VIOLATIONS & REGULATORY CONCERNS - Dwight Chamberlain,

j.

Deputy Director, Division of Reactor Safety

1

i

4.

LICENSEE PRESENTATION -

!

{

5.

BREAK (10-MINUTE NRC CAUCUS IF NECESSARY)

i

j-

6.

RESUMPTION OF CONFERENCE

7.

CLOSING REMARKS - LICENSEE

8.

CLOSING REMARKS - Ellis Merschoff, Regional Administrator

,

,,-,-,, _- ,

, . - -

. - . - , . - ,

, , , , _ . . . . , . - - .

. . , - .

.

-

-. .

.

.-

..

. - . -

.

-

_ . . .

. - - . .

.

- -_ .

..

,

.

,

.

f

i

'

,

[

APPARENT VIOLATION *

i

!

PREDECISIONAL ENFORCEMENT CONFERENCE

!.

!

.

OMAHA PUBLIC POWER DISTRICT

e

[ JULY 21,1997]

.

1.

i

  • NOTE: THE APPARENT VIOLA TION DISCUSSED A T THIS PREDECISIONAL

ENFORCEMENT CONFERENCE IS SUBJECT TO FURTHER REVIEW AND MA Y BE REVISED

PRIOR TO ANY RESULTING ENFORCEMENT ACTION.

!

!

l

j

4

J

,v

.

-

_

.

.

-

e

f

.

e

APPARENT VIOLATION

1.

10 CFR 50.65(a)(1) states, in part, that each holder of a license to operate a nuclear plant

shall monitor the performance of structures, systems, or components, against licensee-

'

established goals, in a manner sufficient to provide reasonable assurance that such

j

structures, systems and components, as defined in paragraph (b), are capable of fulfilling

their intended functions. Such goals shall be established commensurate with safety and,

'

where practical, take into account industry-wide operating experience.

10 CFR 50.65(b) states, in part, that the scope of the monitoring program specified in

paragraph (a)(1) shallinclude safety related and nonsafety related structures, systems, and

components as follows: (2) Nonsafety related structures, systems, or components: (iii)

Whose failure could cause a reactor scram or actuation of a safety-related system,

l

10 CFR 50.65(a)(2) states, in part, that monitoring as specified in paragraph (a)(1) is not

required where it has been demonstrated that the performance or condition of a structure,

system or component is being effectively controlled through the performance of

appropriate preventive maintenance, such that the structure, system or component remains

,

capable of performing its intended function.

Contrary to the above, as of April 21, '.997, for certain nonsafety related structures within

the scope of this rule, the licensee had neither monitored the performance of these

structures against licensee-established goals, nor demonstrated that the performance or

condition of these structures was being effectively controlled through appropriate

'

preventive maintenance such that the structures remained capable of performing their

intended functions. Specifically, the large radius piping elbows of the fourth stage

extraction steam system, sixth stage extraction steam system piping and other piping in

the heater drains system were neither monitored nor effectively controllad through

j

preventive maintenance such that these piping locations remained capable of performing

i

their intended function. This was evidenced by: 1) the second downstream large radius

piping elbow in the fourth stage extraction steam system failed catastrophically on April

21,1997, resulting in a plant transient; and 2) the following piping structures were

subsequently determined to be below minimum wall thickness: a) the furthest downstream

large radius piping elbow in the fourth stage extraction steam system line (S-32); b) a sixth

stage extraction steam system " pup" piece (S-54); and c) three parallel lines in the heater

drains system (D-95).

THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

BE REVISED

Retrieved from "https://kanterella.com/w/index.php?title=ML20149L208&oldid=5491807"