ML20059B313
| ML20059B313 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 12/16/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20059B306 | List: |
| References | |
| 50-331-93-21-EC, EA-93-255, NUDOCS 9401040077 | |
| Download: ML20059B313 (36) | |
See also: IR 05000331/1993021
Text
,_ - . .- - .
4- 0
'
.;
i
U.-S. NUCLEAR REGULATORY COMMISSION i
4
REGION III
Report No. 50-331/93021(DRP) EA-93-255
Docket No. 50-331
License No. DPR-49
Licensee: Iowa Electric Light and Power Company ?
IE Towers '
, P. O. Box 351
Cedar Rapids, IA 52406
Meeting Conducted: November 19, 1993
Meeting Location: Region III Office ;
799 Roosevelt Road
Glen Ellyn, IL 60137
, Type of Meeting: Enforcement Conference
Inspection Conducted: Duane Arnold Energy Center .
August 24 through October 14, 1993 !
Inspectors: J. Hopkins "
C. Miller
Approved By: 'd (2\Nh1 .r
R. D. Lanksbu'ry, Chief', Date
Reactor Projects Sedt4dn 3B
Meetino Summary
Enforcement Conference on November 19. 1993 (Report No. 50-331/93021(DRP) i
Areas Discussed: One apparent violation with two examples identified during-
the inspection was discussed along with the corrective actions taken or i
planned by the-licensee. This. apparent violation involved the inability of
the *B" standby diesel generator (SBDG) to automatically supply.~ power to .
essential bus IA4 had offsite power been lost. In the first example, the "B"
SBDG was inoperable for a time in excess of that. allowed by technical :
specifications. In the second example, core alterations were in progress
while both the "A" and "B" SBDGs were inoperable contrary to technical.
specifications. ,
!
i
i
9401040077 931219 . l
PDR ADOCK 05000331 :
G PDR, ~l
i
. . . ..
_
i
.
f
,
DETAILS
- !
1. Persons Present at the Enforcement Conference ,
i
Iowa Electric Liaht and Power Company
J. Franz, Jr., Vice President-Nuclear
D. Wilson, Plant Superintendent, Nuclear ;
M. McDermott, Manager, Engineering ,
K. Peveler, Manager, Corporate Quality Assurance ,
J. Bjorseth, Maintenance Superintendent !
T. Gordon, Supervisor, Electrical Maintenance -i
T. Allen, Senior Discipline Engineer, Electrical- l
L. Heckert, Principle Licensing Specialists 1
U. S. Nuclear Reaulatory Commission l
i
H. Miller, Deputy Regional Administrator, Region III
E. Greenman, Director, Division of Reactor Projects, Region Ill ;
R. DeFayette, Director, Enforcement and Investigation Coordination l
Staff, Region III ;
L. Greger, Chief, Reactor Projects Branch 3, Region III :
B. Berson, Regional Counsel, Region III
R. Pulsifer, Project Manager, Nuclear Reactor Regulation (NRR) .
J. Beall, Office of Enforcement, NRR (via telephone)
J. Hopkins, Senior Resident Inspector, RIII
Z. Falevits, Electrical Inspector, Region III ,
M. Khanna, Reactor Engineer, Region III j
2. Enforcement Conference j
An enforcement conference was held in the NRC Region III Office cn
November 19, 1993. This conference was conducted as a result of the i
'
findings of inspection conducted from August 24 through October 14,
1993, in which two examples of an apparent violation of NRC regulations I
were identified. Inspection report (IR) 50-331/93015(DRP), dated
November 5, 1993, documented the results of the inspection.
The purpose of. this conference was to discuss the apparent violation, j
root causes, contributing factors, and the licensee's corrective i
actions. During the enforcement conference, the licensee presented the
event investigation, safety significance, causes, and corrective.
actions.
.
The licensee's presentation contained no significant additions to the
description of the event documented'in IR 50-331/93015(DRP).
In addition to the corrective actions documented in IR 50-
331/93015(DRP), the licensee's presentation indicated that the following i
additional corrective actions had been taken:
2
l
l
l
P
'
..
I
- A description of the event was placed on the " Notepad" network for
the Institute of Nuclear Power Operations and on the nuclear plant
reliability data system.
,
1
! * The licensee attended a workshop sponsored by the Nuclear
Maintenance Application Center, a subgroup of Electrical Power
Research Institute, to review the event and draft a revision to
i EPRI's 4160 Vac circuit breaker maintenance procedure
recommendations.
l
- Event specific train'ig was completed for all plant operating i
crews.
l' * All maintenance electricians were trained on the details of the
event and the corrective actions in place to prevent recurrence.
- The licensee determined that the issue was not reportable under
10 CFR Part 21. However, the licensee committed to review its
initial evaluation.
,
- Labels were attached to the 4160 Vac circuit breakers stating that
the correct breaker plunger was to be verified when a circuit
3
I '
breaker was " racked in."
- Licensee event report 50-331/93-008 was issued on October 18,
1993, which described the event. !
l
- The licensee presented the results of its risk assessment of the I
"B" SBDG being inoperable during power operations and while core I
alterations were in progress. The licensee concluded that even
though the core damage frequency was slightly increased, prior ,
training, outage risk management, and existing plant procedures i
minimized the overall risk to the plant and the public. l
In addition to the corrective actions documented in IR 50- !
331/93015(DRP), the licensee's presentation indicated that the following
additional corrective actions were planned or under consideration ,
following the "B" SBDG being inoperable for a time in excess of that l
allowed by technical specifications: '
- The circuit breaker vendor, General Electric (GE), planned to
issue a Service Advisory Letter (SAL) describing the issue'.
- The licensee contracted with GE to review the revised 4160 Vac ,
circuit breaker maintenance procedure. It planned to wait until !
GC completed the evaluation before the revision was issued. .
1
1_ A copy of the licensee's and NRC's presentations are attached to this
i. report. ,
1
l Attachments: As stated >
3
i
..
,
.
-
U.S. NRC REGION lli
DUANE ARNOLD
ENFORCEMENT CONFERENCE
November 19,1993 l
'l
10:00 A.M. (CST) ;
,
E A 93-255
REPOPT NUMBER 50-331/93015
.
'
REGION 11! OFFICE
-
,
799 ROOSEVELT ROAD, BUILDING 4
GLEN ELLYN, ILLINOIS'
t
,
. -
.
DUANE ARNOLD i
, ENFORCEMENT CONFERENCE
Agenda
INTRODUCTION AND OPENING REMARKS: 1
l Edward G. Greenman, Director, Division of Reactor Projects
1
'
NRC OVERVIEW:
l Edward G. Greenman, Director, Division of Reactor Projects
l
SUMMARY OF EVENTS:
Jay A. Hopkins, Senior Resident inspector, Duane Arnold
'
Diesel Generator inoperability
SUMMARY OF APPARENT VIOLATIONS:
Jay A. Hopkins, Senior Resident inspector, Duane Arnold
LICENSEE PRESENTATION AND DISCUSSION: i
Iowa Electric Light and Power Company l
{
NRC FOLLOWUP QUESTIONS l
l 1
CLOSING RE% ARKS: ,
Hubert J. Millet; :Egion 1 Deputy Administrator !
/
'
l
l
,
'
l
.
.
.
APPARENT VIOLATION
,
Duane Arnold Energy Center technical specification 3.5.G.1
states that with one standby diesel generator (SBDG)
inoperable, continued reactor operation is permissible for the i
next 7 days unless the SBDG is made operable. If that ;
condition is not met, an orderly shutdown shall be conducted !
and the reactor shall be taken to hot shutdown within the next.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and taken to cold shutdown within the following 24 '
hours. Technical specification 3.9.D.1. requires that with core
alterations in progress, one SBDG be operable with its
associated standby gas system train and its main control room
ventilation standby filter unit subsystem. If that condition can
not be met, core alterations are not permitted.
CONTRARY TO THE ABOVE
i
a. On July 21,1993, with the reactor operating at approximatel
75 percent power, the "B" SBDG became inoperable. The "B"
SBDG was not restored to operable status within 7 days and
the reactor was not taken to hot shutdown within the next 12 '
hours, or cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,
l
b. From August 7 to 11,1993, with core alterations in progress,
the "A" and "B" SBDGs were inoperable.
l
l
i
The apparent violatio.ns discussed in this enforcement conference are
subject to further review and may be subject to change prior to any 1
resulting enforcement action.
, . - - .
.
,
..
,
"B" SBDG INOPERABILITY TIMELINE i
04/16/92 Completion of " LOOP /LOCA" STP on "B" SBDG.
i
r
07/21/93 Routine maintenance on breaker 1 A401.
07/29/93 Reactor in Hot Shutdown.
\
.07/31/93 Reactor in Cold Shutdown
}
08/04/93 "A" SBDG tagged out for maintenance.
08/07/93 Core Offload Started.
.
08/11/93 Core Offload Completed.
08/22/93 "A" SBDG operable.
,
09/16/93 1 A411 failed to close during "B" side LOOP /LOCA.
.
09/22/93 1 A411 failed to close during "B" side LOOP /LOCA
09/23/93 1 A401 plunger gap adjusted.
,
, 09/24/93 "B" side Loop-LOCA successfully completed 3rd attempt.
09/25/93 "B" SBDG declared operable
L
09/27/93 All auxiliary contacts verified operable "as-found".
.
~
i
09/28/93 Measurements taken on all 4160 breakers. 26 breakers
outside acceptance criteria.
i
,
10/02/93 10 breakers were adjusted.
Retests were cc.mpleted satisfactori,1y.
10/02-04 5 non-essential breakers adjusted. !
10/04/93 Reactor S/U.
i
,
9
.
.
i
i
ENFORCEMENT CONFERENCE
.
?
DUAXE ARNOLD EXERGY CESTER
FRIDAY, NOVEMBER 19,1993
.
d
i
j
I
l
l
1
l
.<
!
'
- l
1
,
5
l
i
T
'
1
.
l
l
DUANE ARNOLD ENERGY CE:5TER
i
l
AGE:SDA
OPENING REMARKS - JOHN FRANZ
VICE PRESIDENT-NUCLEAR
DISCUSSION OF EVENT -
TOM GORDON
SUPERVISOR,
ELECTRICAL MAINTENANCE
&
SAFETY SIGNIFICANCE -
TIM ALLEN
SENIOR DISCIPLINE ENGINEER,
ELECTRICAL
CLOSING REMARKS -
DAVE WILSON
PLANT SUPERIN'ENDENT
I
20/ EMBER 19,1PF3
. . . . . .
.
-
.
b
DUANE ARNOLD E:SERGY CE3TER :
.
J
$
a
DISCUSSION OF EVENT :
i
i
TOM GORDON 1
SUPERVISOR, ELECTRICAL MAINrENANCE- !
,
'!
!
!
3
NOVNMR19,1999
!
l
'l
I
.- . -- -.
.
'
.
DUANE ARNOLD ENERGY CENTER
DISCUSSION OF EVENT
e SEQUENCE OF EVENTS
- Initial LOOP /LOCA Test Identification of Failure
- LOOP /LOCA Surveillance (Figure 1)
- Problem Summary
- ProblemIsolation
- 1A411 Circuit Description (Figure 2)
- Troubleshooting Sequence
- Root Cause Identification
- Successful LOOP /LOCA Test
e CORRECTIVE ACTIONS
- Short Term
- Maintenance History Review
- VendorInput
- Existing Breaker Conditions
- Breaker Restrictions
- Training :
- Industry Notifications
- Long Term
- Procedure Revision '
l
- Industry Notification
- Training -onc
l
NCNEMBER19,19P3
. . . . . . - . . . -. .. - - - . . . . - .
!
.- :
,
9
.i
!
.
i
,-
.
"B" SBDG .
O :
.
) 1'A411 1A4 l
'
4.16 Kv.
? 1 A401 ) 1A402 Essential Buse !
uv STBY us S/U
.1
m Feed -
nn Feed '
i
3
l
-l
I
Figure 1- 1
4160 V ESSENTIAL BUS 1 A4 l
.)
>
-
-l
u
-
. .
. . - .= . . . - .
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
1A411 1C08
k '3d
- IC388O C* y C* R ,.
IA411 1C380 '
Cl* O lC. c c' 4 I{ }I :4
43 /E
c 14
" " " ' #" I
r 10 4
C5
9
Ci - 4r -_152-411E
" :: 143-4 :
CLOSE
.- t43-4
1A40i o
11A401 1 A401 1 A411 4C
-
0-+-
6
,th O h7 : 4
~~
CS o
o
o o
"
186-4 186 o
p
1A411 e ICOS o 7
DG2 o __152-411 CL3 -t152-401
' 1 01 0 1 AOl o 8
b
~
~ 43 /E ' "
o CL3
16r 3
o ; _4152-401 1C118 o IA401 o
152-411 y- "
_
-- b CA ERA
4
33 9 ,
--
o IA402 O
1A401 0 CC --
f 1 o 25-2I o
"
I IIA _%2< IC118 o 7
I A s , Cl2 -t152-402
27-R - ~
25 -
~
T I 3
_ t152-402 1A411 0 8
"
D
27-12: 1- "
b
' " "
__ _ __ _ __ _ __ __ __] :
TDC 7
o o 2
' 5 o gi127-4X
~152-411
- 33
gi127-4X p --
7 o
2 6 _1_5_9
IA411 g 062y
" "
IC388 1A411 l
- 1A402 0
15T 5 1A402 o
-o CL c 0 0 OC N L10- " " " ,
43 /N IC388 1C08 yj 3 35 ;3,43; g
IA43I ~ ~
l
l :: g
. l2 4 16 1 5
. o o o 102-411
I I ! 10PU
l ! 6 Is ~
o
_
FIGURE 2:
DIESEL GENERATOR OUTPUT BREAKER
CONTROL SCHEMATIC
_ - _ _ - _ _ - _ - _ _ - __ - _ _ .
_ _ .
.
-
. t
,
9
1
4
%
1
!
'
Levels
i
!fi?4*ff8'ff*?1
Breaker Plunger Bott >-i / *
f<
square l
2 11 7/32'-11 11/32'
- -
Front (r\;-
!
Lifting rafi ;
Magne Blast 1
4.16 W Breaker
,
9
t
Breaker Lifting Rail-to-Plunger Bolt Measurement l
'
FIGURE 3
,
.;
'l
,
. '
.__
.
.
4160 VAC Magneblast Breaker Plunger Measurements
-(Acceptance Value of 117/32 to 11 11/32 inches)
Plunger toteft Lifting Rail Plunger to Right Lifting Rail
Team A - 11 12/32 11 2/32
Team B . 11 6/32 11 8/32
Team C - 11 6/32 11 8/32
,
O
lt?9*9ff??tf!ti
Breaker Plunger Bolt >> # *
f<
square l
- 11 7/32'-11 11/32'
- :
- . I
O v
{.
,
Ufting rail
Magne-Blast
4.16 W Breaker
,
E
Breaker Lifting Rail-to-Plunger Bolt Measurement
FIGURE 4
!
1
.
Primary Stab to i
Cubicle Rossette / N j
'
-
X interface [ i
N
i
!n
EA x /
/ .- N
\
i
l I N x ' l
l I . ch,-
-
!
'
s
N. .tW
'i i i FT'i 1 l '
"
k 4 l ' I ' I D..
\ ' yJ
\
\
. i
.41
n
1
'
m 'N
-
fx l
/ -~
"za3/32'-1/8"
-
\
__
%%g _, ,____ .%, _p
.
1* - 1 1/8' Travel
I
I !. I I
11 7/32*-11 11/32'
, .I _ =
_
, %M ,
_ m ~~
4.1s w meaner 4.te W Brooker
FIGURE 5: BREAKER DIMENSIONS
- . - - -
,
-
. FIGURE 6
Closec Position >
1
9
=
'
I I ,
'
g w e:,w asyg m aus,
' '
s'A
, ,
'r : i
h) / N k
'
$ h
l, _/ i,t i
,
i
- .
.
2:
s c -;
fj / i
'
- / # i
/ f:g
'
=
$ '
,
5 - -
t
4: u
5 :g ;
-
, r
Y li j i !
l -
Operating
-
f ;
'
Rod ;
d@ 1 ,
i
h !
n
b Operteig Rod l
$-
/ ccm \ .
@ r A
y
1
- t !e e.eund +
q ndn.15N6*
Operating Rod ;' , A.un ,
Casing }j " " * * '
l'
a
s I ,I .
t
h 922 e x:K - i / v;&:wsb/2 s
- ' /ehund .i
0/4' mar) i
A.we i
V )
O/4* b SM) -
- -
. i
Circuit Breaker '
Pksger -l
t
-!
I
r - , -
..
,
i
.
~
FIGURE 7 '
Breaker Open Position :
I l
, . , ~ + . , ;;, , -
I Ik
,l. N
@ ,-
'
.
9 .3 ;
W i
,',
sy )d1
-
w contact
i!j
g
j 33
>
w @(
e
-
ig 1
!
h- _
, ,
'
.y- #, ,
& y if
u
{ Operating- g q
g - -
'B' contact 3
Rod ^
!
-!
(! ;
I
i
Operating Rod R
4 Opermung Rod
Casing s
+
-
u! e \
- . ; -/ ,$
$
n
-
!
l
Vendor manual specified gap
%% r O'-1/8"
_ NwBon ! '
_
w-weetwo
i
!
1
!
I
i
.
Essential 4160 V Magne-blast Circuit Breakers
Breaker Description Adj. Needed Adjusted Comments -
1A301 Standby Transformer Feeder X X
1A302 Startup Transformer Feeder
1A303 480 V load center transformer 1X-31
1A304 Core Spray Pump 1P-211 A X X
1A307 RHR Service Water Pump 1P-22A
1A308 RHR Service Water Pump 1P-22C
1A309 General Service Water Pump 1P-89A
1A310 CRD Feed Pump IP-209A
1A311 SBDG 1G-31 Feeder X X
1A312 480 V Load Center Traasformer 1X-91 X Contacts feed computer point only.
1A401 Standby Transformer Feeder
1A402 Startup Transformer Feeder
1A403 480 V load center transformer IX-41
1A404 Core Spray Pump 1P-2118 X X
1A405 RHR Pump IP-229B X X
1A407 RHR Service Water Pump 1P-228 X X
1A408 RHR Service Water Pump 1P-22D X X
1A409 General Service Water Pump 1P-898 X No Safety related functions (GSW)
1A411 SBDG-1G-21 Feeder X X
l1A412 480 V Load Center Transformer 1X-20 X Contacts feed computer alarm only.
Recirculation Pump Trip Breakers
1A501 Recire. Pump 1P-201 A Trip Breaker A
1A502 Recire. Pump 1P-201B Trip Breaker A X Contacts have no function.
1AS01 Recirc. Pump 1P-201 A Trip Breaker B
1A602 Recire. Pump 1P-201 B Trip Breaker B
FIGURE 8: EVALUATION MATRIX FOR
AFFECTED BREAKERS
_ _ _ _ _ _ _ _ - - . . . - . . - . . . . .- . . ._ -
.
.
Non-essential 4160 V Magne-blast Circuit Breakers
Breaker Description Adj. Needed Adjusted Comments -
1A101 Auxiliary Transformer Feeder
1A102 Startup Transformer Feeder X Evaluated by Systems Engineering
1A103 Reactor Feed Pump 1P-1 A X X
1A104 Reactor Recire. MG set 1G-201 A X Non Critical Plant impact
1A105 Circulating Water Pump 1P-4A
1A106 Condensate Pump 1P-8A X X
1A107 480 V load center transformer IX-11
1A108 480 V load center transformer 1X 71
1A109 480 V load center transformer 1X-51
1A110 480 V switchyard transformer
1A201 Auxiliary Transformer Feeder X Evaluated by Systems Engineering
Startup Trsnsformer Feeder X Non Critical Plant impact
3202
IA203 Reactor Feed Pump 1P-1B
1A204 Reactor Recire. MG set 1G-2010 X Non Critical Plant impact
1A205 Circulating Water Pump 1P-48
1A206 Condensate Pump 1P 88 X X
1A207 480 V load center transformer 1X-21 X Non Critical Plant impact
1A208 480 V load center transformer IX-81 X Non Critical Plant impact
1A209 480 V load center transformer 1X-61
1A210 General Service Water Pump 1P-89C
1A211 Well Water Pump 1P-58D X X
-
FIGURE 8: EVALUATION MATRIX FOR
AFFECTED BREAKERS (Continued)
- - _ _ - . . - - - - . .. . - - ._ _
_ _ _ . . _ _ _ _
Breaker Inspection Section .
TASK CKTBKR-G080-002 Rev.10 GEK-7320F <5000 cycles EPRI NP-7410 Aug. 93
step # allowance step # allowance step # allowance
____ . .__
1 Record initial breaker Operation Counter reading. 5.1.1 x
2 Remove breaker, verify safe and remove barrier 5.1.2-3 a.1
3 Inspect interruptors: 5.1.5 sat /unsat a.3_ det.
damage to the arc chute sides a a.3 det.
~
contamination to the throat area b a.3 _
det.
breaks in the plastisal covering for pole pieces c a.3 det.
4 Check condition of arcing and main contacts 5.1.6 sat /unsat b.2 b.3 dot.
5 Check all bolts are tight 5.1.7
6 Clean breaker 5.1.8 gen.
7 Examine moving joint on load carrying members 5.1.9 galling (y/n) !
8 Examine fiberglass sheets 5.1.10 sat /unsat
5 Lubricate as required 5.1.11 b.7 lub.
10 inspect control wiring and terminations 5.1.12 b.8
~ ii Megger form the studs on Sec. Cpir. to frame 250v 5.1.13 x elec. > = 1.25Mohm
12 Arcing contact wipe 5.1.14 > = 5/16" b.4.1.a > = 5/16" mech. > = 5/16"
13 Primayy,, contact wipe ___ _ . _ __ 5.1.15_ 1/4";5/16"_ _ __ b 4.1.b_ 1/_4"-5/16"_ mech 1/4"-5/16"
_
14 Buffer block clearance to contact arm 5.1.15.1 > = 1/16" b.5 inspect buffer blocks
.]j ndry joliac[ gap __
5.1.16 3-5/8"-3-15/16" b.4.1.c 3-5/8"-3-15/16" mech. 3-3/5"-3-15/16"
16 Trip latch wipe 5.1.17 3/16"-1/4" adj 3/16"-1/4" mech. 3/16"-1/4"
16.1 Trip latch tension mech. 25 lbs.
,17 Trip armature travel _ _
5.1.18 1/16"-3/16" b.4.1.d 1/16*-3/16" (1/32" otvil mech. 1/16"-3/16"(1/32o1)
_
18 Release latch wipo 5.1.10 3 /16"-1/4" b.4.1.f 3/16"-1/4" { mech, no tol. given
19 Release latch monitoring switch 5.1.20 < = 1/3 2 " b.4.1.g < = 1/3 2 " mech. < = 1/3 2"
20 Motor and relay switches 5.1.21 < = 1/32" b.4.1.h < = 1/32" mech. < = 1/32"
21 Interlock switch 5.1.22 < = 1/32" b.4.1.i < = 1/32" mech. < = 1/32"
21.1 C,ontro_t relay contact inspection b.6
22 Oriving a_ndhtching pawl adjustment 5.1.23 > = 0.015" b.4.1.) > = 0.015" mech. > = 0.015"
23 Crankshaf t endplay 5.1.24 < = 0.015" adj. < = 0.015" moch. < = 0.015"
_
24 Prop pin cicarance 5.1.24.b > = 0.025" adj. > = 0.02 5" mech. > = 0.025
25 Latch checking switch 5.1.25 b.4.1.k mech.
contact check 5.1. 25.e < - 1/16" _
1/16" mech. app.1/16"
switch arm gap 5.1.25.g > = 1/64" adj. > = 1/64" mech. > = 1/16"
26 Plunger interlock-measured lif ting rail to plunger 5.1.26 11-7/32"-11-11/32" b.4.1.1 11-7/32"-11-11/32" mech. 11-7/32"-11-11/32"
27 Operating mechanism inspection 5.1.27 b.5 g/det, w/ interlocks
28 Trip coil and armature inspection 5.1.28 sat /unsat b.4.7
29 Low resistance (ductor) readings on each phase 5.1.29 x eloc. < = 54microchm
70v trip test 5.1.31 70v Ctrl Pwr 70v trip, 90v close
_ 30
31 Charging motor check 5.1.32 lyrush 1/4" min. 6.19 brush 1/4" min.
_ ___
FIGURE 9: PROCEDURE REVIEW MATRIX
_ _ -_ -_. _
32 Arc-qnch cylinder, counter, aux.sw. plung:r check 5.1.34 set /unsat g/dzt.
32.1 Manual / electrical stroke test b.9 gen. w/intsrlocks .
3') St iker p atn inspection
,
5.1.35 sat / unset _
34 Aux. switch contact check / inspection 5.1.36 g/det.
~
35 Reassemble 5.1.39
36 Hypot Sky phase-frame, phase-phase 5.1.41 x elec. > = 1 Mohm/kV
36.1 Opening / closing speed test op.cl. test 14 fps cl.,15 fps op.
'37 Ending counter reading _
,
5.1.43 x
38 Secondary coupler check 5.1.44 0.038" gap det.
39 Primary stab inspection det.
~ dU insulation inspection det.
li Prop spring damage check
- - - .
Cubicle inspection Section
. . _ . . _.
TASK cKT8KR-G080 002 Rev.10 GEH-1802X EPRI NP 7410 Aug. 93
step # allowance step # allowance step # allowance
___ _ _ _ _
...__ _____._._.__ . . . _ _ . _
..1 Clean cubicle 5.2.1 m.1
1.1 Resistance to ground and phase-phase m.2
? Fuse holder contact clip gap 5.2.3 grense method
3 Aux, switch mech. vert. and secure 5.2.5
3.1 CI'ock anchor bolts and control wiring m.5
4 Wire connections on aux contacts 5.2.6 sat /unsat _
_ S. Breaker position switch operator check 5.2.7 sat /unsat
G Breaker lif ting moch. inspect and lube 5.2.8 m.3 dot. anti-croop pingr?
7 High voltage bushing check 5.2.10
8 Load terminal inspection 5.2.12 m.4
9 CT secondary termination check 5.2.14 1
10 Wire edge gaurd check 5.2.15 sat /unsat
11 Protective boot check 5.2.16
12 Local handswitch check 5.2.17
13 Interference block to interlock mechanism check oper. 1/1 G"- 1/8" mech. 1/16~-1/8"
14 Breaker lif ting rail to upper stop clearance oper. 3/32"-1/8"
15 Primary stab wipe oper. mech.
start < = 1/8" from top of ball mech. < = 1/8" from top
Iength 3/4"-7/8" mech. 3/4"-1 -1/32"
_
16 Spring discharge cam interlock oper. locked w/bkr < 1/4" up
17 Positive interlock motor clutch gap check oper. > = 1/16'
..
18 Lif ting bracket play mech. 1/16"-1/8"
,19 Secondary disconnect seated mech.
20 Lifting cradle width check bt notch) 21-1/4"-21-3/4"
FIGURE 9: PROCEDURE REVIEW MATRIX (Continued)
. . -
._______-_-_-__ ___- -
. .. ..
.. . .. ..
... , . .
Acceptance Requirements
TASK cxTaxR-o080-002 Rav.10 GEH-1802X EPRI NP-7410 Aug. 93
step # allowance step # allowance step # a!!owance r
Breaker and cubicle clean and tools removed 6.0.1
_1 Positive interlock roller check G.O.3 in V Ond 1/16" gap oper. centered in *V"
2 Plunger / aux. switch gap 6.0.6 some to 1/8" oper. 0*-1/8*
Note: all specs given for 1200 A breaker model AM-4.16-350-2H only
i
l
l FIGURE 9: PROCEDURE REVIEW MATRIX '
(Continued)
l
l
_
i
'
_._m__ ____ _
..
,
-
.
!
DUANE ARNOLD ENERGY CE:NTER
1
l
l
SAFETY SIGNIFICANCE :
1
i
i
TIM ALLEN
i
SENIOR DISCIPLINE ENGINEER, ELECTRICAL
i
t
i
.
!
I
NCNEMBER19,1993
- - .- . . - ._ . . .
.
l
l >
-
j; . ,
DUANE ARNOLD ENERGY. CENTER. i
SAFETY SIGNIFICANCE :
EVENT TIMEFRAMES i
L * ON-LINE EVENT (JULY 21,1993 TO JULY 31,1993) ;
,
'
- B' Diesel Generator Inoperable :
l
- Plant was not placed in cold shutdown within i
required time limits. <
l
l
l * CORE OFF-LOAD EVENT (AUG 7,1993 TO AUG 11,1993) !
'
1
I - Both Diesel Generators Inoperable
- Fuel Movement (Core Off-Load) In Progress ;
l
.
!
DAEC 1
NCn/ EMBER 19,1993 !
!
I
l
_
.
ONLINE DIESEL AUTOMATIC TRANSFER INOPERABILITY .
-
- -
-
-.
. .
.
- .
l
'
- ACTUAL -
"
l ll' ll .
-- -
-
--
,.
-
. . euwr..
.
.
.: .
'
S I
-
- -
- . ; ;
'
.
- : :
- : :
-
!
-
! -
!
, -
.
l unas ;
' =
-
'
TECH SPEC
'-
^ Yr$
. .' .
LIMIT- -W-
-
-
, / / l i
,
-
i ?""""-
.
, .
-
{
. : -
. .
. . . .
,
-
-
^.
i
., : i .
-
- .
DATE:7/21 - 7/22 7/23 7/24 7/25 7/26 7/27 7/28 7/29 - 7/30 7/31 = 8/1
-__-m___m._-+-m-_-_m .-m_r_
w--- _---3w+. -s,wmw% - + _- ws_, -,- si,-e-+ -r r,.s -----s, ,* _ -~e*+- -er,w-=--w*s---==..-- w *ew,w.,-e,-- w-==--ww.+4en--e,-1...-%- m- ---m-a -+.-w ,o
- . .- . .
i
-. ,
i
DUANE ARNOLD ENERGY CENTER -
l
ONLINE EVENT 1
!
l
e Cold shutdown achieved on 7/31/93 exceeded 1
Tech Spec requirements by 1.2 days.
e The diesel generators are required durmg power: ,
operations for a Loss of Offsite Power (LOOP). Diesel !
generator will automatically supply power to its j
associated essential bus.
e Inoperable diesel as allowed by Tech Specs increases j
j
~
the core: damage frequency by approximately 8 fold. .
i
e Exceeding the 7 day Tech Spec limit by 1.2 days increases .
the probability of core damage by an additional j
factor of17%. i
,
e Probabilityis within the bounds of NUREG 880. 1
e Conservatively assumes the 'B' diesel generator cannot !
be returned to service. -
!
'
!
,
'
NOVEMBER 19.1993
i
.
. . . - . - . -
,
-.
!
,
1
!
DUANE ARNOLD ENERGY CENTER l
SAFETY SIGNIFICANCE !
- ON-LINE EVENT (JULY 21,1993 TO JULY 31,1993) ;
i
- The diesel was capable of an automatic start :
'
(rated speed and voltage within the allowed time).
- Diesel generator inoperability was due to. inability to ;
automatically reenergize the essential bus on a Loss of ;
offsite power.
'
,
,
- Output breaker could be manually closed from the i
control room. i
- A loss of the 1A4 bus would have been readily }
recognizable by the operators. t
t
- Existing procedures provide the necessary guidance to ;
reenergize the essential bus.
- A blind test was run on two operating crews. Both crews
identified the failure and reenergized the bus within 5-
l
minutes.
- Event-specific training has been completed for all !
operating crews.
- 1992 licensed operator continuing training speci5cally
addressed 4160v distribution and the diesel generator
,
j
systems. This included the transfer permissives between
offsite power and the diesels.
,
- The 'A' diesel generator was capable of supplying 100%
of the emergencyloads required under design basis 1
accident conditions. l
sov SEEt9,im
. . . _ . . . _.
,
-
,
-
3
l
-
DUANE ARNOLD ENERGY CENTER .
ONLINE EVENT I
.'l
e The calculated effect on plant safety even without :
operator action was small. ;
e Operators had been trained on how to identify and handle i
the loss of an essential bus prior to the event.
,
e Simulator tests demonstrated the operating crews' ;
, ability to promptly restore power using existing , lant '
procedures.
e The 'A' diesel generator was operable and capable of- !
- supplying 100% of the emergency loads required under !
design basis accident conditions.
e
,
i
!
l
. ;
.
-
NOVEMBER 19,1995
l
.
..
.
DUANE ARNOLD ENERGY CENTER
CORE OFFLOAD EVENT
e Fuel offload commenced on 8/7/93 and was completed on
8/11/93: a total of 4.25 days.
e Tech Specs require one offsite power source, and one
diesel generator to be operable during fuel movement.
e 'A' diesel was out of service for planned maintenance.
e 'B' diesel was available with the exception of the capability .
of the diesel generator's output breaker to automatically
close in on the essential bus.
e The probability of a fuel handling accident concurrent
with a LOOP during the 4.25 days of fuel movement
(while assuming both diesels inoperable) is 6 x 10E-8.
This assumes no operator action.
e This probabilityis remote and does not significantly add
to the overall probability of a release.
e Secondary containment was intact and would isolate
on a LOOP.
- The essential bus could have been reenergized from the
Control Room with minimal operator action.
DAEC l
l
NOVEMBER 19.1PP3
i
$
4
-
.
?
DUANE ARNOLD ENERGY CENTER
CORE OFFLOAD EVENT
e Prior to RFO12 an independant shutdown risk review
of the outage schedule was performed.
e The risk factor during fuel move was only 4.5% of the
Phase 1 risk
planned main (even with the 'A' diesel inoperable for
tenance).
e With the 'B' diesel inoperable and no operator actions,
the relative risk during this time frame is increased to
only 4.83% of the highest risk phase of the outage (Phase 1).
e The increase is minimal due to the aredetermined plant
configuration during this phase of the outage:
- all 'B' side safety systems were considered protected
- the reactor cavity was flooded
- two fuel pool cooling systems were availaable
- both SBGT trains were operable
- both S/U and S/B transformers were available
- work in the switchyard was prohibited
e Due to preplanning, the LOOP scenerio only accounts
for approximately 2% of the risk during this phase.
NOVEMBER 19,1993
_ _ _
.
i
-
.. ;
,
!
!
DUANE ARNOLD ENERGY CENTER :
!
CORE OFFLOAD EVENT !
.
- Operators were well trained and kept wellinformed of-
current plant status.
- Briefed on the outage risk analysis study prior
to the outage.
- A daily safety assessment analysis was complete. ;
- Electrical systems were maintained in the " acceptable
risk" category. ;
- Daily time to boil calculations after a totalloss of l
decay heat removal were performed (average tim.e
to boil 35-40 hours). :
o
- The 'B' diesel was not operable, but was available and
could have been closed onto the bus from the Control Ro
l
P
'- J
'
I
DAEC ~
NOVEMBER 19,19PS -
'!
1
..
,
t.
DUANE ARNOLD ENERGY CENTER
'
CORE OFFLOAD EVENT
s The probability of a concurrent LOOP and fuel handling
accident is low.
e The second inoperable dieselincreases the probability
of boiling to less than 5% of other acceptable phases
of the outage.
- Operators were trained on the outage risk study prior
to RFO12.
'
e Operations was kept informed of daily outage risks including
equipment out of service and time to boil.
- Outage preplanning and defense in depth minimized the
effect of the 'B' diesel inoperability.
e The diesel can be considered available and operator
actions taken into account.
)
,
,
wac
, NOVEMBER 19,1995
l
I
l
. - . . . .
-
.. :
t
,
!
DUANE ARNOLD ENERGY CENTER
.
SAFETY SIGNIFICANCE CONCLUSIONS '
!
!
- Iowa Electric fully recognizes the importance of :
maintaining our diesels in an . operable status. o
e The inability of the 'B' Diesel to automatically ;
supply power to its essential bus had a small-
i
effect on risk even without operator actions.
l
e This failure was analyzed prior to the event and is l
included in both the online and RFO12 risk l
analysis. 1
u
e The operators had been trained on how to ;
recover from the event and.were well informe'd i
of the current status of plant safety systems and .
time to boil during RFO12.
- RFO12 risk assessment and preplanning
minimized the effect of both diesels being inop
during core offload using the defense in depth
philosophy.
!
l
4;
I
'
Y
ji
!)
-
NOVEMBER 19,1995
i
!
_. .. .
o -. , 3
DUANE ARNOLD ENERGY CEMER >
i
I
f
&
CLOSING REMARKS- !
.
i
!
>
p
DAVE WILSON !
PLANT SUPERISTENDENT !
,
f
1
'
,
t
.l
.
.
.
DAEC 7
NOVEMBER 19.1993 .
,
h
1
-
.. . _ .. - _.-
. .. ;
,
DUANE ARNOLD ENERGY CENTER- :
MI G TI G FACTOR
!
e SELFIDENTIFIED
.
1
e NO PRIOR OPPORTUNITY TO IDENTIFY j
o PROMPTREPORTING
i
'
e COMPREHENSIVE SHORT-TERM
CORRECnVE ACTIONS !
-
e ISOLATED OCCURRENCE
1
e STRONG SUPPORT OFINDUSTRY
NOTIFICATION
e PRIOR TRAINING, RISK MANAGEMENT:
AND EXISTING PROCEDURES MINIMIZED 1
EFFECTS ON PLANTSAFETY l
.!
i
.
. .
l
-1u ,
tue j
.l
'l
l
.. . - - . . - - - . - . ..
- :. .
DUANE ARNOLD EXERGY CEMER
ATmNDEES:
JOHN FRANZ -VICE PRESIDENT-NUCLEAR
DAVE WILSON -PLANT SUPERINTENDENT
MIKE MCDERMOTT-MANAGER, ENGINEERING
KEN PEVELER -MANAGER, QUALITY ASSURANCE
JOHN BJORSETH -MAINTENANCE SUPERINTENDENT
TOM GORDON -SUPERVISOR,
ELECTRICAL MAINTENANCE
TIM ALLEN -SENIOR DISCIPLINE ENGINEER,
ELECTRICAL
LARRY HECKERT -PRINCIPLE LICENSING SPECIAIlST
,
NOVEMBER 19,1993