ML20245D071
| ML20245D071 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Sequoyah |
| Issue date: | 05/17/1985 |
| From: | Jacqwan Walker TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML082490787 | List: |
| References | |
| AOI-14, NUDOCS 8711040348 | |
| Download: ML20245D071 (149) | |
Text
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Sequoyah Nuclear Plant DISTRIBUTION 1C Plant Master File s
1U Regulatory Engg Supv (4) SE Training Office IC Unit 1 Control Rm IC Unit 2 Control Rm IC SE (+2WC)
IC Operations Supv ABNORMAL OPERATING INSTRUCTION (2),SE Training,0ffice IU Plant bt t (0&E)
AOI-14 1U SE Trainin ffice IC Chief NUC Training B
(+1U)I
)
LOSS OF RHR SHUTDOWN COOLING IC NUC PR Dnc ce r1.
1Mn PRT -
3U Plant Manager, WENP Units 1& 2
_1U NEB, W10C174 C-K IC Resident NRC Inspector - SNP IC NSRS, E7A2-3C-K 1C Technical Support Center IU Compliance Supv IU NSGPO III Class 1U INPO, c/o Lu Yarger, Manager Evaluation Support Dept 1100 Circle 75 Pkwy Suite 1500 Atlanta, GA 30339 1U Technical Services 1
Prepared By:
J. R. Walker R: vised By:
C. A. Jetton
/
Submitted By:
/Aa Q
Superv or PORC Review:
Mg 17 585 Date
'I Approved By:
M NM Plant Manager D:te Approved:
M...w i 7 EB5 j
g l
The curr'ent revision level of this instruction is:
8 Rarson for current revision (include all temporary change numbers) Revised to add section G in table of contents.
Tha last page of thid instruction is number 24 92 1888! Bu8$g7 P
- .7...
q 4
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}
Sequoyah Nuclear Plant.
ABNORMAL OPERATION INSTRUCTION:
f I
Cover Sheet Page 2 AOI-14 History of Revisions
.)
{.
j Rev. No.
Date-Revised Pages Rev. No.
Date Revised Pages l
3 7/25/79 ALL
).
4-6/24/81 3, Add 5 &6
)
5
'3/14/83.
A11~
i 6
07/05/84
'All p.
7
_10/12/84-lAll l
i 8
05/17/85 2
1
~
l l
1 4
l-i 4
1
...__________--J
SQNP AOI-14 Unit 1 or 2
)
Page 1 of 24 Rev. 7 LOSS OF RHR ' SHUTDOWN COOLING' e
b I.
SYMPTOMS 5
'1.
Uncontrolled increase in RCS temp due to failure of RHR cooling.
2.
Abnormal RHR system operation.
II.. AUTOMATIC ACTIONS None III. ' IMMEDIATE OPERATOR ACTIONS '
None 7
)
i
). ':
+
SQNP-AOI-14 -Unit 1 or 2 Page 2 of 24 l
Rev. 8 LOSS OF RHR SHUTDOWN COOLING l
).
~IV. - SUBSEQUENT OPERATOR ACTIONS Section A.-
FCV-74-1 or 2 Closed L
Section B -
RHR Pump Trip i
Section 'C I RHR System Leak p
L Section D. ' Loss Of CCS To RHR Section E -
Alternate Heat Sink Using Steam Generators Section F -
Alternate Heat Sink Using Spent Fuel Pool Cooling 4
Section G -
Venting of RHR Pumps 'After Loss of Suction Due to g
Low RCS Water Level When in Modes 5 and 6
{
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SQNP=
AOI-14 Unit 1 or 2 Page 3 of 24 Rev.- 7 LOSS OF RHR SHUTDOWN COOLING' SECTION A FCV-74-1 OR 2 CLOSED J
}
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS l-I 1
Initiate REP Per IP-1 Notify Operations Duty I
a.
Specialist within 5 min 1
1 2
Stop RHR Pumps il 3
Close RHR crosstie Prevents depressurization to CVCS and flashing.in RHR piping a.
Close FCV-62-83 1
4
=
s.___________________ - __ - _
-c SQNP AOI-14 Unit 1 or 2 i
Page-4 of 24 l
~
Rev. 7 i
LOSS OF RHR SHUTDOWN COOLING-
)
i SECTION A - - FCV-74-1 OR 2 CLOSED
{
L
)
l STEP
' ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS l
I l
4 Decrease RCS Press
< 380 psig
/
a.
Decrease press in CAUTION this preferred order i
IF an RCP is running,
- 1) Open pzr normal sprays THEN maintain RCS press
> 325 psig
- 2) Establish aux spray AND seal A P > 200 psid when pzr/ spray A T
< 320*F
- 3) Decrease charging flow while maintaining RCP seal injection > 6 gpm
- 4) Increase letdown with PCV-62-81 and orifice
. valves
- 5) Establish excess letdown
- 6) IF RCS temp > 350*F, TIIEN cooldown RCS using main condenser or S/G PORVs
SQNP AOI-14 Unit 1 or 2 Page 5 of 24 Rev. 7
.i LOSS OF RHR SHUTDOWN COOLING-SECTION A
}.
FCV-74-1 OR 2 CLOSED i
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS o
5 Restore Shutdown Cooling 1
a.
.WHEN RCS press FCV-74-1 and 2 auto
< 380 psig, isolate at 700 psig and
_THEN open FCV-74-1 and 2 cannot bc opened until
< 380 psig b.
Close RHR heat! exchanger outlet valves FCV-74-16 and 28 c.
Close RHR heat exchanger bypass valve FCV-74-32 l
d.
Ensure CCS established to RHR heat exchanger and RHR seal cooler
Control RHR heat exchanger outlet valve and bypass valve as required g.
Control charging and letdown as required h.
Open RIIR crosstle to CVCS FCV-62-83 l
l
- End Of Section A.-
s.j
.i SQNP AOI-14 Unit 1 or 2 s.',,
Page 6 of 24
~
Rev. 7.
LOSS OF RHR SHUTDOWN COOLING-SECTION B RHR PUMP TRIP p-
' STEP-ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS 1
1 Initiate REP' Per'IP-1:
I a.
Notify Operations Duty
'l Specialist within 5 minutes 2-Close RHR Crosstie.
. Prevents depressurization To CVCS
' and flashing in RHR piping a.
Close FCV-62-83 t
3 Restore Shutdown Cooling
~
a.
Ensure FCV-74-1 and 2 IF FCV-74-1 or 2 closed, open TliEN refer to Section A b.
Close RHR' heat exchasger outlet valves FCV-74-16 and 28 c.
Close RHR heat exchanger bypass valve FCV-74-32 d.
Ensure CCS estabushed to RHR heat exchanger and RHR seal cooler e.-
Start the RHR Pum'p IF no RHR Pump avauable, that' did not trip TiiEN refer to Section E l
f.
Control RHR heat exchanger outlet valve and bypass i
valve as required h.
Open RHR crosstie to CVCS FCV-62-83
~.
SQNP AOI-14 Unit 1 or 2
' Page 7 of 24 Rev. 7 '
(
LOSS OF RHR SHUTDOWN COOLING-SECTION B RHR PUMP TRIP STEP
' ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS
.a l
[
4 Evaluate Cause Of RHR Pump Trip
- End Of Section B -
t 6
i s-l i
I l
l l
) !,
l E
SQNP!
l
. AOI-14~ Unit 1 or 2 Page. 8 of 24
{
Rev. 7' i
i LOSS OF RHR SHUTDOWN COOLING-SECTION C RHR SYSTEM LEAK p
I STEP ACTION / EXPECTED - RESPONSE-CAUTIONS / REMARKS i
i 1
Initiate REP Per IP-1 1
l a.
Notify Operations Duty
-l Specialist within 5 minutes
-t e
2 Identify RHR System Leak j
,,o a.
Check aux. bldg area monitor recorders RR-90-1 and 12 b.
Locally inspect aux bldg and entmt i
i c.
Check CCS radiation Identifies leak in RHR heat monitors exchanger and/or RHR pump seal 3
Isolate RHR System Leak IF no RHR train available, TIIEN refer to Section E a.
IF required, place unaffected RHR train in service l
i End Of Secdon C -
h='*
SQNP AOI-14 Unit 1 or 2 1
- Page 9 of.24 Rev. 7
]
LOSS OF RHR' SHUTDOWN COOLING-~
1 SECTION D
1 J
1 STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS i
I.
l 1-Initiate REP Per IP-1 a.
Notify Operations Duty.
i
)
Specialist within 5 minutes 2
- Stop RHR Pumps With Loss Prevents' flashing-in RHR Of CCS -
heat exchanger and/or RHR pump, seal' CAUTION When CCS becomes available, then restore CCS to RHR heat exchangers slowly to reduce thermal shock 3
Close RHR Crosstie Prevents depressurization To CVCS and flashing in RHR piping a.
Close FCV-62-83 1
4
.g.
i
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- SQNP '
' AOI-14 Unit' 1 or 2 -
Page.10 of 24 Rev. 7 LOSS OF RHR SHUTDOWN COOLING-SECTION D - - LOSS OF CCS TO RHR STEP ACTION /EXPEC' TED ' RESPONSE i
CAUTIONS / REMARKS,_
):
4 Restore Shutdown Cooling 4
a.
Ensure FCV-74-1 and IF FCV-74-1 or 2 closed,.
2-open TIIEN refer to Section A '
.b.
Close RHR heat exchanger outlet valves FCV-74-16 and 28 c.
Close.RHR heat exchanger i
bypass valve FCV-74-32 d.
Ensure CCS established to -
RHR heat exchanger.and RHR seal 4
e.
Start R.HR ' Pump IF no RHR Pump available, TIIEN refer to Section E f.
Control RHR heat exchanger i
outlet valve and bypass valve as required g.
Open RHR crosstie to CVCS FCV-62-83
- End Of Section D -
t 1~
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s 1 i
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SQNP j
AOI-14 Unit 1 or 2
{
o Page 11 of 24 Rev. 7 LOSS OF RHR SHUTDOWN COOLING' SECTION E ALTERNATE HEAT SINK USING STEAM GENERATORS h
)
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS 1
Initiate REP Per IP I a.
Notify Operations Duty i
Specialist within 5 min 2 ~
Ensure RCS Pressure IF RCS pressure boundary Boundary Intact TOT intact, I
THfN refer to Section F a.
Reactor vessel head on and tensioned I
b.
UHI connected to vessel head 3
Restore RCS Press Minimum press for RCP
> 325 psig operation I
a.
Control pzr heaters I
as required
)
R I
I q D j
f SQNP AOI-14 Unit 1 or 2 i
Page 12 of 24 l
Rev. 7 '
LOSS OF RHR SHUTDOWN C'OOLING-
{
SECTION E ALTERNATE HEAT SINK USING STEAM GENER ATORS l
. STEP ACTION / EXPECTED RESPO_NSE CAUTIONS / REMARKS j
q 1
4 Isolate RHR System -
a.
Close RHR discharge j
valves
- 1) FCV-74-16 l
i i
- 2) FCV-74-28
- 3) FCV-74-32 j
- 4) FCV-62-83 b.
Stop 'all RHR pumps I
c.
Close FCV-74-1 and 2 5
Start An RCP a.
Loop 2 preferred b.
Refer to SOI-68.2 i
= l
t p
AOI-14 Unit 1 or 2 Page 13 of 24 i
Rev. 7 l
LOSS OF RHR SHUTDOWN COOLING ~
SECTION E ALTERNATE HEAT SINK USING STEAM GENERATORS
}.
s STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS 6
Establish Secondary Heat Sink.
a.
Control S/G narrow range
-(
level between 25% and 50%
j
-)
J b.
Dump steam to condenser IF S/G water subcooled, i
using condenser dumps TliEN feed and bleed S/G J
or S/G PORV using AFW and S/G blowdown 1
l l
i l
- End Of Section E -
l
' l l
SQNP AOI-14 Unit 1 or 2 Page 14 of 24-Rev. 7
)
LOSS OF RHR SIIUTDOWN COOLING-1
{*
- nn-l ALTERNATE HEAT SINK USING SPENT FUEL POOL COOLING SECTION F
...,.s,,
I STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS I
1 l
i NOTE This section should NOT I
be used if normal RHT
~
cooling or secondary heat L.
sink is available 1
Initiate REP Pei' IP-1 a.
Notify Operations Duty Specialist within 5 minutes v
l 4
1
]
SQNP AOI-14 Unit 1 or 2 Page 15 of 24 Rev. 7 l
LOSS OF RHR SHUTDOWN COOLING-l SECTION F ALTERNATE HEAT SINK USING SPENT FUEL POOL COOLING i
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS 2
Establish Flow Path From RCS To Spent Fuel Fool
!f a.
Reactor vessel head off IF reactor vessel head on, THEN ensure UHI grayloc connections removed b.
Blind flange from fuel transfer tube removed c.
Blind flange from refueling cavity to lower entmt installed d.
IF time permits,
' Assuming 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after THEN comp!cte sealing shutdown, RCS temp refueling cavity per
= 140 F, and RCS level FHI-6 at the vessel flange, then i
RCS will start boiling in
- 20 min and RCS level will reach fuel level in s 80 min if the operator took no action e.
WHEN water level in SFP and refueling cavity are
- equal, THEN open fuel transfer tube gate valve f.
Ful refueling cavity with Control injection flove to RllR, SI Pumps, or CCPs maintain RCS subcooled, from RWST to RCS but minimize flow in order to extend the time before SFP cooling is required.
Refer to appendixes A and B. -
SQNP AOI-14 Unit 1 or 2 -
Page 16 of 24 Rev. 7 J
LOSS OF RHR SHUTDOWN COOLING-SECTION F ALTERNATE HEAT SINK USING SPENT FUEL POOL COOLING 1.
)
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS l
3 Prepare To Instau SFP Spool Piece a.
Stop SFP Cooling Pumps b.
Isolate SFP System from spool piece
- 1) Close 78-511 Aux Bldg 714
- 2) Close 78-512 Aux Bldg 714 l
- 3) Close 78-513 Aux Bldg 714 l
- 4) Close 78-599 Aux Bldg 669
- 5) Open 78-597 to Aux Bldg 690 drain line
- 6) After SFP line drained, then close l
78-597 j
1
- 7) Open vent 78-591 Aux Bldg 714 l
i i
. l i
~.
SQNP l
AOI-14 Unit 1 or 2 Page 17 of 24 Rev. 7 -
LOSS OF RHR SHUTDOWN COOLING-SECTION F ALTERNATE HEAT SINK USING SPENT FUEL POOL COOLING-STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS I
i 4
Prepare To Install'RHR Spool Piece
- a. - Isolate RHR System from spool piece (valves are unit specific)
- 1) Close FCV-74-33
- 2) Close FCV-74-35
- 3) Close FCV-02-83
- 4) Close HCV-74-34 RHR Hx Rm B and 74-534
.s
- 5) Close HCV-74-37 RHR Hx Rm B and 74-529
- 6) Close HCV-74-36 RHR Hx Rm A l
I and 74-528 l
b.
Drain RHR to spool piece i
i
- 1) Open vent 74-539 RHR Hx Rm A
- 2) Open drain 74-540 RHR Hx Rm A O
s -_____-__ -_
l'
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)
' AOI 14 Unit 1 or 2 Page 18 of 24 ~
t-
.. ~
Rev. 7 e
LOSS OF RHR SHUTDOWN COOLING-
)
l SECTION Fi - - ALTERNATE' HEAT SINK USING SPENT, FUEL POOL COO' ING.
L I
STEP ACTION / EXPECTED RESPONSE
[ CAUTIONS / REMARKS i
.)
r{
' c~
y y
/
f Spool Pieces
/3 I:
I' ' /J o
a.
Refer to FPMI-12.23
,T 6
' Fill' And Vent Flood Mode Cooling Line
/
f Close drain 74-540' a.
c
/
- b. ' Open 78-513 to refill line CAUTION >,
'i Minimize leaktige from vent valves W539 and 78-591
>o 1
1 c.
Slowly open HCV-74-36
)
g '.
(
.j or 37 to refill line
+
/
t q
J d.
WHEN lines vented, 1
/
THEN close 74-539 and
/
l T~B91 e.
WHEN lines filled (flow
/
j into SFP),
'j i
THEN close 78-513 9
4 f.
Open SFP heat exchanger 0
\\;
outlet valves78-511 and 512 n
-\\\\.
- e; 4
4
-18
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- I,yM,A...g <.
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a-t z!pgqyp./; g N
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'AOI-14; Unit 1 or. 2 aii l
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q'%g Page 19 of 24
.f']
l.
g Rev. 7 2
j 4
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' LOSS OF RHR SHUTDOWN COQ.lg(f 1,,
.l 23 n-
+
7 SECTION 'F
- ALTERNATE HFAT SINK USING SP,ENT Tk{L POOL' COOLING y j,
\\lf "gf
\\
's
^
STEP ACTION / EXPECTED ' RESPONSE CAUTIONS / REMARKS ;
)
ii.
t \\,
-y 7'
j' '
3, a.
Stop RNRJtunps:
b iv-4i
- b.. Close. RHR ! discharge valves 74-S20 and 521 i
q
. T' e
g A'
c.
Close RHRicrosstie -
i valvas FCV-74-33 and 35 b
\\s
-d.
Open RHB,iniection valves
(
i FCV-63-9U and 94
>f e.
Open RHR, crosstie valves I gi J
.1) HCV-74-36 and 528
\\
g 1
-)
- 2) HCV-74-37 and 529 I
- y' q
i 1
f.
Start SFPlacoling pumps
,~
as required y[,,
1
{
t
\\
g.
Control RHR hea't' exchanger 1(.
S" i
~
,1 outlet valves FCV-74-16 i /,
and 28 as requirdd 3
r.
)
^l l
Ebd ' f Section F -
1 O
v i
1
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- l
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m- - --
, q' q
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- ?m y
- %J : SQNP.
gr < 3, m.
.,, _ 4 f
. A01-14 : Unit 1 or. 2 '
- ?. ;g,
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-l 78Ee 20.of 24 g..
7.;
Rev. 7c.
- ij#
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- LOSS'OFjt_HR SHUTDOWN COOLING-
.j j'v s vt a - p; a c
..SECTION G ;- WF,NTING OF Rl!R PUMPS AFTER LOSS OF SUCTION DUE TO.
I LOWRCS. WATER LEVEL WHEN IN MODES 5 AND 6 '
t-
,y ptt c
y-c..
.s.
' " CAUTION:
.OBTAIN HP. ASSIS1'ANCE FOR MONITORING RADIATION BEFORE l' "jj 9 QENTING PUMP.
no y
N SjFf4 hdTION/ EXPECTED' RESPONSE CAUTIONS / REMARKS '
> Y<,
.1
-/
Verify RHR pump (s) A-A 4(B-B)2 switch in " PULL' TO ~,
ec
% % LOCK". position.-
\\'
7 2
/ -
Unlock and close pump dis-charge valve HCV-74-520 (521)2, j
~
[ - Unlock and.open pump dis-3
/
i cfinrge vent HCV-74-516 (517)2, vent pump and piping for.10 mhiutes.
S i
i
'4
/'
(1)Close and relock HCV-74-516-.
(517)2- -
(2)Open,1 5'
'/
and relock 'HCV-74-520 (521)2 4
6
/
(1)dhse RHR Hx bypass valve HCV-7S-36 (37)2 1
4 7
/
)())Close manualtisolation crosstie y'
,6 tier RHR Hx A '(B)2 to CVCS o
J HCS-74-500 (531)2
,.y a
s 8>
1/ "
\\
(1)Close RHR Hx A (B)2 bypass valve FCV-74-33 (35)2
.)
9
/
~
i t 3'(1 + 4)2 FCV-63-93 (94)2 1
i,m 1
4(1) Double person verification required per OSLA 58 att, D U( )2. Indicates train IO equipment -
(
r 4
f k.
i e
i
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o e
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SQNP AOI-14 Unit 1 or 2 Page 21 of 24
-l Rev. 7 i
LOSS OF RHR SHUTDOWN COOLING' SECTION G VENTING OF RHR PUMPS AFTER LOSS OF SUCTION DUE TO LOW RCS WATER LEVEL WHEN IN MODES 5 AND 6 CAUTION:
OBTAIN HP ASSISTANCE FOR MONITORING RADIATION BEFORE VENTING PUMP.
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS 10
/
(1) Open RHR Hx A (B)2 NOTE: ALLOW PRESS. ON outlet FCV-74-16 (28)2 pI-74-13 (26)2 TO GO TO ZERO, THEN LEAVE OPEN FOR 5 MINUTES.
11
/
(1)Close FCV-74-16 (28) 2 CAUTION: MAINTAIN CONSTAN1 LEVEL SURVEILLANCE ON RCS
' LEVEL SIGHT GLASS DURING PUMP RUN PERIOD.
IF LEVEL DROPS RAPIDLY, IMMEDIATELY j
STOP RHR PUMP.
IF A SLOW E
LEVEL DROP RESULTS IN EX-HAUSTING VCT VOLUME TO LESS THAN 20%, OPEN FCV-l 62-135 TO MAKEUP, FOR hi SHORT PERIOD, OBSERVING LEVEL INCREASE CLOSELY ON RCS LEVEL SIGHT GLASS TO PREVENT OVERFILLING RCS.
I 12
/
(2) Start RHR pump A-A (B-B)2, NOTE:
IF OBTAINING A verify mini-flow FCV-74-12(24)2 SOLID A SOLID SYSTEM IS u
opens. Observe amps, flow and NOT ACCOMPLISHED ON pressure on pump for erratic FIRST ATTEMPT STOP RHR behavior.
PUMP (s) AND REPEAT STEPS 1-5 OF PROCEDURE TO VENT PUMP EASING AND PIPING AGAIN CONSULT SRO BEFORE RUNNING PUMP (s) AGAIN.
l 7
13
/
(2) Slowly open FCV-74-16 (28)2 CAUTION: DO' NOT EXCEED f
. to discharge any remaining gases 3500 GPM TOTAL FROM BOTH to RCS.
PUMPS TO RCS TO PREVENT VORTEXING AT RHR LETDOWN LINE ON #4 HOTLEG 1
(1)
Double person verification required per OSLA 58 app. D.
l
( )2 Indicates train B equipment
~
SQNP AOI-14 Unit 1 or 2 Page 22 of 24 Rev. 7 l,
LOSS OF RHR SHUTDOWN COOLING-L.
SECTION G - VENTING OF RHR PUMPS AFTER LOSS'OF SUCTION DUE TO l
LOW RCS WATER LEVEL WHEN IN MODES 5 AND 6
- CAUTION:
OBTAIN HP ASSISTANCE FOR MONITORING RADIATION BEFORE VENTING PUMP.
STEP ACTION / EXPECTED RESPONSE CAUTIONS / REMARKS i.
14
/
After gasses are all exhausted from system, as observed. by normal constant pump amps, i
and pressure, STOP A-A (B-B)2 RHR pump and return to normal standby cooling configuration.
L L
l.
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r i
( )2 Indicates train B equipment
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SQNP F
i n.._._.... %
._...i.............
AOI-14 Unit 1 or 2
.-.. -. j"*.
l i
Page 23 of 24 i
L
-a Rev. 7
._..__p.._._...._...,...
-Appendix A
{
i I
- 4. - - - -
FLdW_ RATE REQUIRED To MNNTAIN RCS.TEMPERATURtCO%TAhlT.; n.....
5 i
1 i
t__._..
I i
40co:
- y. -
ME-A,BER-Trip '
}..._.
4,
.......a l
[
)
T a.
s 35%
' e-, TEW-140*5 l
4
! 5".-TUA. Pd8#_1.-
=......
4
'3
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ATTACHMENT 6 TO ENCLOSURE 1 s
81051740566 UN11F:D STATF;M (;oVF;ItNM);NT 66667N26 ff 7 Yemorandum TENNESSEE VALLEY AUTHORITY
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T PORC Chairman and Members, Sequoyah Nuclear Plant
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3 0 *18 J. H. Sullivan, Supervisor PORS, SB-2, Sequoyah Nuclear Plant p.gg7 n4TF:
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SEQUOYAH NUCLEAR PLANT (SQN) - INVESTIGATION OF kATER SPILLS FROM STEAM GENERATOR MANWAYS OCUP' DING JANUARY 28 and February 1,1987 Attached is the subject report for your review in accordance with Technical Specification 6.5.1.6 (g).
I plan to present this at the normal PORC meeting on Apell 7, 1987 for PORC's joint review and to answer any further questions.
J. H. Sullivan
.N JHS:RAF.
Attachment cc (Attachment):
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Rit1S_.MR AN 72 A-C, /
NRC Resident, O&PS-2, Sequoyah NER Coordinator (Landy McCormick), O&PS-4, Sequoyah Ben Lake, Operations Training PORC Secretary, Susie Holloway O&PS-2. Sequoyah
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BBQUOYAH NUCLEAR PLANT paep.,,.
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UNIT 1 1
l INVESTIGATION OF WATER SPILLS j
FROM STEAM GENERATOR MANWAYS OCCURRING JANUARY 28 AND FEBRUARY 1, 1987 1
MARCH 198*1 i
ASSESSMENT TEAM MEMBERS
,/77 kIM
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\\ J. M. Alexander, Jr., Shift Engr, Operations, SQN D. A. Barker, Mech Engr, Mech Test, SON
./ <
t%vG.o2A G. R. Donneau, Asst. Shift Engr, Site QA, SQN 8
S. K. Chapman, Support General Foreman, Elec. Mtn, SQN h
B. D. Childs, Shift Supv, Radiological Control, SQN l
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_11. H. Gammage, Mech Engr, Plant Assessment, SQN
//
R. M. Hodge, Mech, Engr, S/G Group, SQN V. M. Taylor, Safety Specialist, Safety, SON
. [dv L
J. S. Woods, Supervisor, PORS, VBN t
_/
' ASSESSMENT T LESDER IA I ///7 J. H. Sullivan, Supervisor, PORS, SQN l
t/
c PORC Review Date PORC Chairman PORC comments / actions derived from this report:
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CONTENTS
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Objective 6
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II. ' Scope
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7' III. Executive Summary 8
IV.. Background 9
V.
Chronology of Events 10 A. General ~
10 B.. January 28, 1987 Event 10-12 C. February 1, 1987 Event 13-14 VI.
Findings 15 N.
A. January 28, 1987 Event 15-17
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N B. February 1, 198'l Event 17-26 o.
VII. Recommendations 27 A. January 28, 1987. Event 27 B. February 1, 1987 Event 27-29 C. Peripheral Issues 29 VIII. Conclusions 30 A. January 28, 1987 Event 30 m.
B. February 1, 1987 Event 31-32 IX.
References 33-35 X.
Attachments 36 l
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NOMENCLATURE l
Assistant Shift Engineer ASE 700 Assistant Unit Operator CB Boron Concentration Combustion Engineering, Incorporated CE Cold Leg CL
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Chemical and Volume Control System CVCS C-Zone - Contamin'ation Zone ECW
-Engineering Change Notice Electrical Maintenance EM Eastern Standard Time EST Plow Control Valve FCV b.
Gallons Per Minute GPM HCI
- Hazard Control Instruction HEPA
- High Efficiency Particulate Air Pilter CN HL
- Hot Leg Hold Order HO HP
- Radiological Control C
IM Instrument Maintenance Institute of Nuclear Power Operations INPO 77 Licensee Event Report LER Main Control Room MCR MOVATS - Motor Operated Valve Analysis Test System Maintenance Request MR Mechanical Test Section MTS MV
- Manway
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OSLA Operations Section Latter Administrative Public' Address PA PkT
- Plant Operations Review Committee PORS' Plant Operations nevtew Staff, PPM Parts Per Million PRO Potential Reportable Ocdurrance PRS Plant Reporting Section PT Pressure Transmitter PVSCC Primary Water Stress Corrosion Cracking
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QA Quality Assurance QMDS Qualified Maintenance Data System RB
- Reactor Building RCS Reactor Coolant System RKR Residual Heat Removal c3 RM Radiation Monitor RWST
- Refueling Vater Storage Tank SE
- Shift Engineer r=
SER Significant Event Report e,.
S/G Steam Generator e.
S/G TA - Steam Generator Technical Advisor S/G PM - Steam Generator Program Manager SI Surveillance Instruction SNP
- Sequoyah Nuclear Plant SOER
- Significant Operating Event Report l
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SQA Standard Practice SQN-Sequoyah Nuclear Plant TC Temporary Change TI Technical Instruction TVA Tennessee Valley 1.utitority UBHT U-Bend Heat Trace U0 Unit Operator VCT Volume Control Tank M
-Westinghouse Electric Corporation VR Work Request F G O
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OBJECTIVE The objective of the Assessment Team's investigation is'to identify the
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facts surrounding the steam generator manway water spill incidents of January 28, 1987 and February 1, 1987.' The initial investigation wil.1 center around the second event with the objective of identifying actions' necessary to resume work safely. The root cause of the water spills will be determined. Immedia'.a corrective action taken will be evaluated and "
further corrective action identified / recommended., Action to prevent acurrence will be recommended and the above will be documented in this report and presented to the Plant Operation Review Committee (PORC) for consideration in accordance with Technical Specification 6.5.1.6.g.
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e II. SCOPE The scope'of the investigation and resulting report is to cover the events inanediately before, during, and af ter the water spills of January 28, 1987 and February.1.,,1987. The resulting findings, conclusions.,,and recommendations will be limited to what was learned from the
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investigation. Peripheral issues that are identified during this investigation may be listed as items needing further research, but this report is not intended to bring resolution to those item s.
This report will be provided to the Pxperience Review Coordinator for appropriate dissemination.
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' !!I.- EXECUTIVE SLNmARY on January 28, and February 1,;1987, approximately 500 gallons and 3,000 i
gallons of water, respectively, spilled from the primary manways on the
~l unit 1 s/Gs which were open for row I and 2 tube heat treatment and eddy current testing. Other than the fact that water was spilled, the events have little in common. Both events illustrated the need for prompt effective communications between the MCR operators and the U/G workers.
Prescriptive communication was established prior to restarting S/G work' i
after the February 1, 1987 event. The first event'contains greater nuclear safety ramifications and the second event contains greater industrial personnel safety ramifications.
^
The January 28, 1987 event resulted from the lack of an adequate RCS level indicating system. The single indicating system malfunctioned and caused actions which lead to a low RCS level, oscillating 1A,RHR motor current, reduced pump flow, RHR pump stoppage, loss of shutdown decay heat removal, fo11oded by a high level and a flow of RCS water out of the open SG manways, onto the reactor building floor." A diverse level. indicating system is recommended. This event was reported to the NRC as LER 50-327/86012.
The February 1, 1987 event resulted from an inadequate procedure for stroking a valve that was at one time an acceptable instruction. The procedure was adequate for the normal conditions when the RCS is pressurized'to greater than or equal to 30 psig but is inadequate for the case where the RCS is less than 30 psig. Contributing to this event was an operator error 641en the procedure was realized to be. inadequate.
Supplemental steps were taken without~ going through the change approval process or referring to flow diagrana. This resulted in gravity flow of water from the RWST to.the RCS and the resultant spill. Personnel injury Ch was avoided by the timing of the event. Personnel responsible for the
, preparation, performance, and management of any part of the surveillance program must realize the significance of their procedures, the need ibt adequate procedures, and the requirements for stopping and obteining an approval change to an inadequate procedure before proceeding.
- r-Additionally, a method to ensure important information is not deleted from a procedure during the revision process must be implemented. This event I'
was reported to the NRC as LER 50-327/87013.
m Following the February 1, 1987 event, interim measures were recommended by the task force (Attachment 11) to strengthen the lines of communications and the caution orders associated with the makeup valves from the RWST.
The subject valves will also have their breakers opened whenever someone is.inside a S/G bowl.
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I IV. BACKOROUND OF STEAM OENERATOR WORK L
During steam generator work in the September 1985 unit I cycle 3 refueling outage-(U1C3 RPO), the decision was made to preventively plug all row I tubes in the Sequoyah unit 1 S/Os to prevent a forced outage (s) during the nest operating cycle fron* row 1 Primary Water Stress Corrosion Cracking L
(PWSCC). Subsequently, the unit was not returned to service because of the Environmental Qualification program _and other work. Up until approximately July 1986, the schedule was not confirmed to determine if U-Bend Heat Treatment (USHT) could be implemented.* The UBHT process had not been qualified and deninstrated in September 1985. During the third l
quarter.of 1986, the propos.t1 to unplug.the row 1 tubes and perform the I
UBHT was presented to TVA management and the decision made to proceed to '
do the work in the current extended outage.
A contract was awarded to Westinghouse (W) Electric Corporation for the USHT portien of the work and a contract was awarded to combustion l
Engineering, Inc. (CE) for tube plug removal, for use of the CE CENESIS manipulators during UBRT, and for subsequent eddy current testing and replugging. Both contracts were awarded on December 12, 1986. During the l-week of December 15, 1986, the CE and W equipment were mated for mock-up I
testing at SQN to identify potential problem, areas _.
A decision was made by the S/G PM and his staff not to use nozzle dams 1 -
during the UBHT since a flood-up of the reactor cavity was not necessary.
(
Because of bolting problems associated with usage of nozzle dams at other utilities and the man-rem exposure for installation, removal, etc., it is not desirable to use the nozzle dams _unless necessary for refueling operations. During previous outages at SQN, S/G work had been performed without any type of RCS overflow problems.
CF The level indicating system at Sequoyah consists of a permanently installed sight glass in fan room number one connected to the bottom of the crossover leg of Loop 1.
A TV camera is utilized to transmit the level information to the MCR.
This system was installed due to the reliability problems SQN h~ad experienced'with the make-shift tygon tubing.
e the st'am generator work had Up until the January 28, 1987 event, e
proceeded without any indication of problems with the Reactor Coolant System (RCS) water level. One exception to this was a call from Operations on 1/21/87 notifying the S/G workers of a loss of safety injection block on the RCS.
An evacuation was ordered from around the open S/cs but the safety injection block was reinstated without incident.
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Chronology of Events' 3
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A.
CENERAL EhII/I151-
--12/ 26/ 8 6-*
S/0 draining completed
)
1/3/87 RCS drained to allow S/G manway removal 1/3/87 RCS Tagged on Hold Order 1084
- 1/5/87 S/G #1 primary manways removed 1/5/87
, S/0 #4 primary manways removed
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1/5/87' Combustion Engineering on site activities started 1/6/87 S/G #2 primary manways removed 1/6/87 S/G #3 primary manways removed 1/10/87.
Operations notified deplugging work to start (WR) 1/19/87 Deplugging of the row 1 hot leg tubes in S/0 #2, 3 and #4 and both cold and hot leg tubes in S/G #1 were completed 1/19/87 Westinghouse onsite activities started.,
1/23/87 operations notified that UBHT work to start 1/27/87 First actual production heat treatment performed.
- s B. January 28, 1987 Event
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1/28/87 SI-673 in progress which documents RCS level from sightglass indication each 30 minutes All items in eastern standard time.
0033 Made up to RCS by increasing VCT level from 28% to 56%. Makeup was from RWST.
20 gal /\\.
Approximately 560 gallons total.
0100 SI-673 reading indicated RCS level normal 0120R adC
0130 00 could not see top of' water column in sightglass using monitor. Sucpected lighting in el fan room may have changed
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which can make it difficult to see top of water column. Sent ADO to inspect' lighting'and sightglass. SI-673 Jeadinge were
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not taken during the period in which the level could not be determined. ADO found sightglass indication 11" above last reading and completely off the scale in use. He verified vent line open and verified that the sightglass scale could not have been :noved. This level indication (696'6") was logged in the SI-673 time slots which had not been recorded earlier.
0330 Not being able to explain the level increase but assuming it to be correct, the Do began to decrease level back to the normal level of 695'6".
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0620 Stopped 1A-A RHR pump when amps and flow rate began to
- fluctuate, miniflow valve operated.
Entered LCo 3.4.1.4
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. action. Operator referred,to A01-14 and SOI-74.1A.
'Sightglass indication was at 696'4".
The 00 decided that the level indication was faulty. Do began increasing level in. preparation to start IB-B RHR pump. Sent AUOs to vent 1A-A RHR to insure no air was in the pump. Sent ADO's to
,5 align la heat exchanger bypass flow and letdown. Without letdown through RHR, RCS level was increasing from charging flow and level increase was noted on sightglass.
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0625 operations called S/G Technical Adviser and requested that S/G platforms be, clear,ed.
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0630 S/G Technical Adviser notified HP and personnel were removed to the laydown area Cs 0714 00 stopped charging pump and closed or verified closed VCT
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outlet valves and RWST supply valves when notification was received from CE trailer that water level was near spillover level in S/cs. Sightglass level indication was at 696'10".
e Water spilled over for a reported 10-15 minutes while IB RHR C'
letdown was being aligned.
f3 0730 RCS water level stabilized by operations. Conversation between SE and PRS Supervisor determined that the IB-B RHR pump was operable and, therefore, no 10 CPR 50.72 reporting requirements applied.
1-PCV-63-1 could be opened to the header to operate the 1B-B pump.
0745 An extra UO entered containment. iiolated, drained, and reopened valve 1-68-551 to the sightglass. Level returned to initial level at 697 as read locally.
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0750 LCo 3.4.1.4 action exit.
1A-A RHR pump venting complete per AOI-14.
1B-B 7HR pump in, operation.
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0803 operations ope 5ed the'.480-volt bfeakers.for lower containment
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0815 Team of TVA, contractor personnel, maintenance personnel, and HP sent in to secure equipment to a dry area and to assess damage.
Ap' proximately.1/2" of water was in laydown area and
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puddled on the platforms.
1000 All electrical connections were disconnected from power service (breakers were already open) by electricians and contractors instructed that they would be responsible to test equipment and reconnect as needed.
1130
-CE GENESIS Consultant notified that laborer clean up complete
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3 allowed to enter for equipment maintenance.
1200 y given permission to start s,et-up.of equipment.
1230 Electrical, Maintenance sent in to check power supplies and to reconnect electrical power.
1530 S/c #3 UBHT equipment recalibrates'and started production heat treatment.
1600 S/c #2 UBHT e.quipment recalibrates and ready for restart of
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work.
Os 1740 Operations ~had tygon tubing hooked up to provide a redun' dant check of RCS level.
1950 Operations and IMs flushed sightglass supply and removed black particle matter and crud.
g 2248 Operations raised RCS level to 695'-8" to verify operation of level indication on sightglass and tygon hose.
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1/29/87 0200 operations reflushed sightglass numerous times. Dirty water and some particles resembling Black Beauty were.. revealed.
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February 1, 1987 Event l
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l'/30/87 operations receive,d work package (SI-166.8) from Mechanicai Test Section to~perford a stroke test on valve 1.-rcV-63-1.
' package received on 1/30/87 with requirement to perform before 2/2/87, 2/1/87 Decision by SE, ASE, and Do to perform SI.
AGE and 00 discussed caution order 1003.
1254 IBB RHR pump stopped ' for SI 166.8 on 1-PCV 63-1.
op'erator recognized the instructions in SI-166.8/166.3 were not adequate for the current plant configuration (S/G manways open), operator closed valves'FCV-63-93/94, rcV-74-16/28/32.
Stroked valve 1-PcV-63-1 (42 seconds total for open and close).
Level in sightglass immediately went oft' ecale high as monitored by the TV camera (Elevation 697' 2").
Water
- observed coming from the manways over the cameras in the S/6 trailers and HP control point.
Individuals inside containment observe water flowing from manways and evacuated to outside polar crane wa!!.
1256 Uo restarted RHR pump to go to maximum let60wn. HP and S/G trailers called Control Room to notify them of water coming from S/G manways.
1257 S/G Program Manager notified of event by S/C Trailers.
1302 Do dispatched AUo to 669 pipe chase to verify 1-PcV-63 1 Closed.
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ADO at valve, verified identification tag number, then hand tightened the valve. Another AUo sent into containment for inspection.
p-1315 S/G pH called to determine what had happened and requested that 480-V power to the RB outlets be turned off. UO confirmed 480-v power off.
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1335 RCS level established at 695'-9".
S/G pH contacted operations Manager at home and discussed incident. operations Manager agreed to come to plant.
1500 Meeting between S/G Group Mechanical and Electrical Maintenance, IIP, and cont'ractor personnel from CE and W.
Decision made to go in and assess damage and condition of nozzle covers.
1618 Placed fio #1156 on 480-V power to RB outlets.
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l 1630 EM and CE enter for inspection. S/G PM and HP Shift Supervisor enter containment for inspection.
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Released HO #1156 and. closed b'reaker to 480-V Ha. outlets.
1825
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1935 S/G PM and HP Shift Supervisor in to replace nozzle covers.
S/G #2 and #4 cold leg nozzle covers were still in place.
1944 S/G #2 HL nozzle cover replaced.
1949 S/G #3 HL nozzle covhr replaced.
1954 S/G #3 CL nozzle cover replaced.
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'2010 S/G #4 HL' nozzle cover replaced.
2015 S/G #1 CL-nozzle cover replaced.
2020 "S/G #1 HL nozzle cover replaced.
2045 Laborers _ enter containment for clean-up.
2145 Laborers complete S/G #2 & #3 platform cleanup. Vorking on laydown area.
2215 HEPA filter orifices removed for full ventilation flow through S/Gs.
2/2/87 c5 0015 Cleanup on'laydown area under S/G #2 and #3 complete.
0245 V enters containment to check equipment damage.
0521 V went to S/G #3 platform and removed heater from R2C57 tube, Heat cycle was. interrupted and heater left in the tube during e-the event.
C' 0830 Plant Manager assigned PORS Supervisor to assemble and head up F?
an investigation team to look into the two recent S/G spills.
S/G work is on hold until management is satisfied it can be done safely.
1030 PORS Supervisor assembled Task Force to initiate investigation.
1200-2300 Vith special permission from the P'lant Manager. CE removed GENESIS manipulators from S/G #2
- 3, and #4 for maintenance.
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VI.FIRDINGS
- k. January 28, 1987 Event 1)
' PRO 1-87-047 was initiated on 1/28/87 to document the event (RCS level malfunction, dralndown, RHR motor 1A-A current oscillation, RHR Pump 1A-A stopped, makeup through charging pump, spillaver from S/G manways of approximately 500 gallons). SQA 186 was coepleted 1/29/87 (Attachment 7).
An initial review by the SE and the Supervisor, plant Reporting Section, deter. mined that the 1B-B pump was operable and therefore, there was no 10 CFR 50.72 notification required.
, 2)
In the early morning of January 28, 1987, approximately 560 gallons of water had been made up to the RCS/VCT.from the RWST.
There was no communication between Operations and the S/G workers prior to or during this makeup.
3) ite Do could not determine RCS level due to his. inability to.
see the sightglass water level. Water level was found 11 inches higher than previous r'eading (approximately 2-1/2 hours earlier).
C-4)
The CO, believing the sightglass reading to be correct, started lowering the' water level. *For about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> the level was being lowered by approximately 30 GPM (letdown minus charging) but the sightglass had only dropped from 696'6" to 696'4".
The, level should have dropped approximately 10' inches. The RRR motor IA-A current started to oscillate (28 1 10 amps) and the U0 stopped the pump. Operations notified S/G Ch personnel to clear the S/G platforms and then they started increasing level, venting 1A-A'RHR pump, and aligning IB-B RHR pump. For approximately 44 minutes water was added t.o the RCS e".
via the charging pump until the sightglass indicated 696'10" and the S/G personnel. notified the control room that the level e
was near sp111over. The UO shutdown the charging pump and isolated the flow paths (VCT and RWST). About three minutes C'
later water started spilling over from the manways and I'
continued for a reported 10-15 minutes. Approximately 30 minutes later, when ADO entered containment, the water level was 697' on the sightglass, and 696'10" on the control room monitor. RRR pump 1B-B was placed in service to reduce level via RRR cleanup to CVCS.
5)
The sightglass was flushe'd, drained, and air was blown through the sightglass. Black particles and crud were observed in the sightglass. Operations initiated a periodic (weekly) flush of the sight glass (SI-673).
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0120R aaC g
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- 6) Boron Concentration Pre-Event 0
~
RCS Temperature Approx. 95 P Sourcet Recorder chart 1-TR-74-14 Attachment 17 RCS CB = 2151ppe Sample: 1/28/87 0005 EST SI-38 required ca = 1987 ppm Source: SI-38 1/27/87 0900 EST Post Event RCS Temperature Apprcx. 1150F Source: Recorder chart
'1-TR-74-14 Attachment 17 RCS CB = 2186 ppe Sample:
1/29/87 0230 EST SI-38 required Ce =.1987 ppm Source: SI-38 1/28/87 1000 EST 7.) U-Bend Heat. Treatment (UBHT) was in progress Ix) both S/G #2 and #3 during the January 28, 1987 RCS spill. As soon as Safety and HP concerns were properly addressed, both CE and W dispatched personnel to the S/G area to move equipment to dry areas and check for water damage. No damage was reported by either contractor even though some control panels and especially the W probe pusher
-~
did get wet.
Due to the slow rise i~n water level, the nozzle covers were not displaced. A check of the equipment compared to each contractor's accountability log showed no items lost due to the flood-up.
o..
- 8) The Health Physics Technicians assigned to the S/G outage monitor the activities of all S/C related personnel. This is done by continuous personnel coverage, video and audio communications 06 systems, or a combination of these. All work is coordinated through HP and the necessary RVPs,, dosimetry, respiratory protection, etc., are issued from the S/G control point. Any full or partial entry into the S/G bowl is directed and timed by v.
personnel at the control point desk in communications'with the worker doing the' entry, the HP on the platform, the S/G control trailer, and the laborer supporting the dressout. RP controls access to the S/G bowls by locking the shield doors (attachment 3) between entries by authorized personnel. Just prior to the 3
incident of January 28, 1987, the DO notified the S/G control trailer of the possibility of RCS spilling out of the manways, thus all personnel were removed from the platforms to the laydown areas. There were communications when RCS started filling the S/G bowls and all personnel were removed from the laydown areas to the raceway. All personnel were in the raceway before RCS began pouring from the platforms. Step-off pads were placed at the polar crane wall doors and dressout requirements were increased for all entries until conditions returned to normal. Exposure for the cleanup operation totaled 0.84 man-rem.
)
- 0120R aac
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1
- 9) on January 29, 1987 at approximately 1820 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.9251e-4 months <br />, the operations Manager gave verbal instructions to the shift engineer to place a note in-the shift engineer and Ro/SRO Log Books about verifying level between Tygon hose and permanent mounted sightglass once per
.4 shift and in all cases at least once per day.
l
)
8.* February 1. 1987 Event
- 1) Flow control va'.ve 1-FCV-63-1 was modified (ECH 6686, 'WP 12150) by disconnecting the brakes and regearing. This modification slowed the stroke time of 1-PCV-63-1 t' rom approximately 15 seconds to -
j approximately 20 secnnds. Work request WR B123699 and
~;
'L SI-166.6/166.3 stroking instructions satisfactorily accomplished the post modification testing for WP 12150 on August 16. 1986.
l The data-package for testing this valve was on the Masterlist and therefore was not forwarded to the hechanical Test Section (Mrs) for review but retained in the control room.-
l
- 2) The scheduled quarterly performance of SI-166.3 was performed on l
November 10, 1986 to stroke 1-PCV-63-1.
The results of.the stroke
~
time (20.3 seconds) mandated the valve to be placed on an increased frequency every 31 days instead of every 92 days. Valve l
1-PCV-63-1 should not have been placed on increased frequency since a modification had been performed to' slow the valve stroke time.
o.
- 3) The events in the main control room occurred as follows. Valve 1-PCV-63-1 was on increased frequency testing. SI-166.8 issued January 31, 1987 due by February 2, 1987. The Uo was going to
. perform SI-166.8/166.3 and record the closing time of 1-PCV-63-1.
The 00 determined the instructions were inadequate and attempted to perform the instructions without going through the instruction
.p change process. The 00 discussed the procedure inadequacy and the L*-
proposed valve alignment with the ASE and SE.
The U0 did not refer to the flow dia' grams. The U0 stopped RHR pump 18-D, closed
~"
valves 1-FCV-63-93, 1-PCV-63-94, 1-PCV-74-16, 1-PCV-74-28, and 1-PCV-74-32 (refer to Attachment 6).
These precautions were taken because the Uo was cognizant of the work on the S/cs and knew the
'RHR system would have to be isolated from the RCS.
The operator was allowed to deenergize the RHR pumps for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance
! P' with the ** part of LCo 3.4.1.4.
The RHR Loops were OPERABLE and the action statement of LCo 3.4.1.4 was not entered. Two valves 1-PCV-74-1 or 1-PCV-74-2, the suction valves to the RHR pumps from the RCS were not closed. This allowed a flow path from the RWST into the RCS.
RCS level, as observed on sightglass with the camera / monitor, immediately went high (off scale) when 1-PCV-63-1 was opened.
1-PCV-63-1 traveled to the full open position and was immediately closed by the Uo using the MCR handswitch. The Uo knew to reduce RCS level. The RHR system was aligned for maximum I
letdown and the RHR pump was restarted. The level did not j
decrease as quickly as the Uo thought it should, therefore the Uo, I
thinking 1-PCV-63-1 was not ! ally closed dispatched an AUo to hand tighten 1-PCV-63-1.
l.
0120R l
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- - o tie U0, ASE, and 7E 'did not refer. to the flow diagrams to confirm I
the valve alignment and RHR system isolation. Valves 1-PCV-74-1
~
and 1-rcV-74-2 are in series and. closing either valve would have isolated the flow path.
~
~
-- 4) Water flowed by gravity from the RWST through 1-PCV-63-1, 1-63-502, 1-PCV-74-2, 1-PCV-74-1 and into RCS hot leg Loop 34.
(refer to Attachnent 6).
The water first flowed out of the hot leg manway of SG 94, then the other hot legs, and finally the Eold legs. The HEPA filter hoses on cold leg manway shield doors (refer to Attachment 4) for S/G 1 and 3 fell due to the weight of the water in them. The water flowed from S/G #4 hot leg manway at a much higher rate. Water flowed out over the top of the door on SG #4 hot leg manway and water was still dripping down out of the tubesheet 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the event.
Eight people were in the react'r. building and six of those were o
5) inside the crane wall. Pour people were in the S/G 2/3 area, tso in the S/G 1/4 area, and two were outside the crane wall. Work was in progress placing the end effector on the manipulator located in steam gene'rator #2.
Communication was being maintained between the RP Control Station S/G. trailers, and the people working on the Steam Generator platforms. People not involved in the work activity in progress were staying in the ALARA tent in os the laydown area (refer to Attachments 1 & 2),
o-Health Physics Technicians or.dered an immediate evacuation of the area when the person working on Steam Generator 2 and 3 platform
="
stated that water was coming from the manway shield doors (refer to Attachments 3 & 4).
All people in the area followed the C6 instructions of the Health Physics Technicians and evacuated the area in an orderly and controlled manner. Since the established step off pads were wet, the Health Physics Technicians moved the step off area to the" raceway of the containment. Approximately one half to one and one half inches of water covered the Reactor Building floor when the water stopped flowing.
g.
C' Health physics and the S/G trailers notified the operators in the Main Control Room that water was coming out of the Steam Generator manways. No announcements were made over the Public Address (PA)
System.
0120R aaC
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6)L A HP technician, a support laborer, and a Westinghouse platform-worker were in the ALARA tent area of 5/o 2/3 in-lower containment at. initiation of.the event (refer'to Attachments 1 & 2)..Another Westinghouse platform worker was ch S/G #2 platform. lHe had.just.
g closed the door to-the hot-leg manway and was about to re-open 1
that door when he heard wrter running. He then saw water coming from s/o #3 and #2 manways.. Water was running off the platforms' at the ladders for both S/c 02 and 83.
He decended'the. ladder at
's/o 32 moving to the back side of the ladder under'the herculite to get out of'the flow of the water. A TVA S/0 Support Engineer and another !(P technician were in the S/0.1/4 area at the time of.
I the event initiation. The support Engineer had been working on s/c e4 platform and was about to return to the platform when he heard what sounded like the ventilation system was starting and then water started. pouring off the S/G 1.6 4 platforms.
The HP-technician had just made it out of the inner C-zone by'way of the step-off pad but the support Engineer could not exit that way due to the water falling from S/G 81-platform. After waiting a short period, he.left the ares under the S/0 #4 platform, six individuals.got wet, none were contaminated. The HP technicians' exhibited exceptional' professionalism in execution of their duties of providing HP support for S/G Work activities.
7)' For the incident on February 1, 1987, there was no warning or prior communications that a problem could occur. Personnel were l
both on the platform and in the.laydown areas. The highest priority was to get all personnel to safety as quickly as possible. With water flowing over the step-off pads for the inner and middle S/G zones and the impending danger of electrical hazards, it was proper to establish a new location in'the raceway for the removal of rainsuits and respirators. Step-off pads were Ch' placed at the polar crane wall doors by the HP technicians prior to them exiting containment. Dressout requirements were increased to include rainsuits"and respirators anywhere inside the polar
(.-'
crane wall for all entries until the cleanup was in progress. The
}'
requirements were adjusted as conditions returned to normal.
1~
Exposure from this cleanup and repair operation totaled 4.82 man-Rem and lasted from February 1, 1987 at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> through Pebruary 6, 1987 at 1830 hours0.0212 days <br />0.508 hours <br />0.00303 weeks <br />6.96315e-4 months <br />.
t:
In both incidents the HP air samplers stopped running when the electrical power outlets were deenergized. Containment air monitors 1-RM-90-106 and 1-RM-90-112 were checked after each event and indicated no appreciable increase in activity. When power was restored, air samples were pulled and confirmed that there was no increase in activity from samples prior to the spills.
Exposures received during the two spills was equivalent to approximately 4.0 days of work by CE, W, and TVA personnel at the average exposure rate of 1.4 man-rem / day prior to these spills.
0120R aac
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- 8) The video cameras inside the S/d lost power at event initiation.
The external camenos trained at.the manways showed a major gush of water for about 3-5 minutes, followed b'J approximately 10 minutes of decreasing flow.
- 9) Inadequate communications, in conjunction with the caution order on 1-PCV-63-1, existed between the MCR operaiors and the S/G workers.
A
- 10) Recovery actions by the Control Room personnel to enssure 1-FCV-63-1 was clos.sd and to decrease level was prompt'And e f f ec t ive.- The operators believed 1-Pcv-63-1 didn't fully close and therefore tightened it by hand. The operators were able to increase letdown by starting t'ne RHR pamp. The findings on the as-found McVATs test of 1-rCV-63-1 are contained in reference 20.
- 11) The Steam Generator program Mansgrr ves notified by the Control Trailer that water was coming from 'the S/G and responded properly to ensure the safety of personnel under his direction.
- 12) The Manager, Operations Grcup responded to the event by coming to the plant (Sunday afternooni. 'He.gove verbol instructions to the Shift Personnel and then issued a night order on maintaining RCS r=
level (refer to Attachment 9).
- 13) PRO 1-87-056 andSQA-186wasinitiht'ed(refertoAttachment8).
The NRC resident inspector was actilled.
- 14) The urgency of stroking 1-PrV-63-1 was communicated by a night order dated January 9, 1986 (sic), (really January 9, 1987) where Cs a $50,000 fine or escalated enforcement action was discussed (refer to Attachment 9).
E 2
- 15) Testing history of 1 PCV-63-1:
r-08/14/86 SI-166.4 performed Stroke time 15.3 seconds C'
08/16/86 SI-166.6 performed Stroke time 20.46 seconds.
I'
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This was the Post Maintenance
]
Test (PMT) for VR B123699 and WP 12150 that changed the gear ratio and disconnected brakes.
Test package was retained in MCR.
10/7/06 SI-166.6 performed Stroke time 20.35 seconds. This was PMT for WP 12090 that removed the opening circuit torque switch. Test package was rer.ained in NCR.
0120R. aac i
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inwas -
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.i 11/10/86. S h26.? performed Stroke time 20.3 seconds. Since Id
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stroke time had. increased more J
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than 25% from previously known, test. the valva was placed
'y/
on increased frequency by Mechanical Test Section.
12/6/86 SI-16F,.8 performed Stroke time 20.3 seconds. Valvt</
tested monthly due to being on
~
increased frequency.
,y 01/2/87 SI-166.8 performed Stroke time 20.3 se'conds, valve tested ne thly due to being on increased freque' icy.
r 02/1/87 SI-166.8 performed Stroke kl.~e U.3 ' seconds. Valve tested monthly due Eo b91ng on increased frequency.i T
.(
J 02/4/87 SI-166.6 performed Stroke time 20.29,secchds.
performed usi' post main,tendra,g TC 87-117 as ae fT s,.
1
- 16) Six nozzle covera were out of position and hv/ to be reinstalled.
and two were still in place. All nozzle covv/r parts were x
inventoried and accodnted for.
s
- 17) The follcuing floor and equipent drain suw/sdilled
?4vels were recorded and utilized to determine the volume of waub The level g'
decrease in the RVST was not noticeable. TI-28 was used to
. convert tnese levels to the equivalent volume o! water:
s
~
Sump t.evel Ecuivalent Volume Volume Added 40%
630 gallons y
85%
1310 gallons 680 gallons
/ ch 8%
190 gallons 1,
j 1 {, }
35%
560 gallons 370 gallons 8%
190 gallons y
80%
1225 ga1lons 1035 ga1lons y
8%
190 gallons 78%
1185 gallons 995 gallons 8%
190 gallons s
Total volume spilled toto containment 3080 gallons 7,
In addition to the water collected in the sump there was also water added s
to the RCS that.did not flow out the manways. Using elevations provided by the Steam Generator Group <this, volume of water was also estimated.
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- 1) RCS level prior to event 695' 8 1/2"
- 2) Elevation of louer edge of manway 697' 2"
- 3) 90% of the volume of hot and' cold leg piping was filled prior to
(
P event.-
- 4) 50% of vessel volume is void a
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Increast,in Vr.sse 1 891 gallons t
1 Increase in Hot L*.'g Piping 274 gallons
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Increase in Cold Leg Piping
'247 gallons
_n /,4,
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Incresse,in Steam Generator 1421 callons
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Total Increase in RCS level 2833 gallons
.),.( 1./ 'I s
Therefore, the total flow through 1-FCV 63-1 is estimated at 5913 gallons.
e 7
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- 18) The, historty '$f. SI-166.,3 changes af fecting FCV-63-1 (refer to
&(s ftthc:vnent 7.0) ;
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,' [ReviQp$_,0 of SI-166.3 specified conditions for testing of
)
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6 (1 FC,J43-1, that the plant is in cold shutdown, and reactor coolant system,qrst)sureisequaltoorgreaterthan30psig. When
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closed. ]pf FCV-63?! is completed, stroking the data sheet leaves valve I
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b, RevisioQ,/
., \\ V of SI-166.3 specified conditions for testing of
{V,e FCV-63 -1/, that the plant is in cold shutdown, and reactor coolant system ressure is equal to or less than 30 psig.
S m
4
~
p pevision 2 of SI-166,3 changes the valve position when stroking of y FCV-63-1 completed from closed to open.
4, Revision 5 of SI-166.3 specified conditions for testing of FCV-63-1, that.the plant is in cold shutdown, and the RlIR system is isolated from the regetor coolant system.
1 Revision 9'of s/-If6 7 specified conditions for testing of FCV-63-1, that r{c. plant is in cold shutdown.
2
~
The requirement for' /C$ pressure to be greater than or equal to 30 i
,/
4.,,
psig was adder'.dLfing informal PORC Review (Revision 0).
On I
t y',
Revision 1 it.bppears that when someone was trying to correct a
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typing error, they went the wrcag way with their greater than sign 1
i and changed the meaning to less than.
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Revision 5 corrected the pressure problem by isolating the RHR system from the !.03.
This is an adequate condition for testing, but does not give a definite lineup to isolate the RHR system from
- RCS, On Revision 9, the isolation of the RKR system from the RCS was dropped. This left the system line up for stroking of,,
~
FCV-63-1 to the operator performing the test. On Revision 2', when the operator was finished with the stroking of PCV-63-1 the' data sheet stated thei "as left position as open".in mode 5 the normal position for Fev-63-1 is closed. This procedure leaves line up" for stroking of valves to the operator assuming that the operator will perform the proper' lineup required for stroking each valve.
In some cases the procedure gives proper valve lineup.
- 19) On the evening of February 4, 1987, 1-PCV-63-1 was tested to l
f deterinine-if any electrical or mechanical degradations had occurred which would have contributed to the water spill that-occurred on February 1, 1987.- 1-PCV-63-1 was last tested by Electrical Maintenance on Augus.t 16, 1986, after Electrical Modifications had remoyed the motor brake and regeared the operator (references Work plan 12150; WR-B123699). The testing was performed to ensure that the valve vendor seating requirements were maintained and to obtain new HOVATS signatures due to the change in stroke times. previous QMDS maintenance and the initial MOVATS saseline signatures were performed on MR-AS48691 in February of 1986. The valve vendor requires a minimum seating thrust of 6,500 pounds to close this valve under the design flow, pressure, and temperature. The maximum allowable seating thrust as allowed by the valve vendor is 32,600 pounds. On August 16,
~~
1986, the seat thrust at torque switch trip was 10,046 pounds with a total of 30,416 pounds. No electrical or mechanical degradations were detected at this time (reference 20, Attachment e
1).
On February 4, 1987, two MOVATS tests were performed. The first test showed a torque switch trip value of 10,396 pounds with a total of 30,378 pounds, and the second test showed a torque switch trip value of 10,493 pounds with a total thrust of 30,087 pounds. No valve or operator abnormalities were detected during the testing (reference 20, Attachments 2 and 3).
Following the j
r;-
last test, the valve was placed in manual to determine if the valve had closed. With the MOVATS equipment still installed, a half handwheel turn was applied providing a thrust value of 35,200 pounds which was above the maximum value recommended by the valve l
vendor. At this time, the manual test was terminated.
.he MOVATS equipruent was removed from the operator, and the valve wat stroked and leak tested by Operations personnel per SI-166.6 (refer to 1 5). Since no degradations were apparent from the l
MOVATS signatures and the valve passed the leak test, no further iq action is necessary on this valve.
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- 20) Boron Concentration Pre-event e,
C RCS Temperature 0
109 F" rource Oo/ASE Daily Journal RCS C B
2150 ppm 2/1/87 0737 BST SI 38 required C3 1987 ppm 2/1/87 0921 EST post Event RCS temperature 1090F source'UO/ASE Daily Journal RCS CD 2154 ppm 2/2/87 0550 EST SI-38 required CB 1987 ppm 2/2/87 0946 EST
- 21) Although no injuries were reported during or immediately after the water spill which occurred on February 1. 1987. the potential for serious injury did exist.
A.
Iniury Due To Initial Surce of Water -
knjury from this incident was prevented only by the timing of the release of water.
Had the release occurred while.an employee was inside the bowl of the steam generator or es when the lead shield doors were removed, a serious personnel injury could have P"
occurred. The force of the water could have knocked the employee from the platform.
B.
Iniury'Due To Fall While Exitino The Area B
After The Incident -
Injury from this incident was prevented by the calm actions of all the involved.
' employees. Descending the platform ladder in a bubble suit is difficult without the added hazard of falling water.
C.
Iniury Due To Electrical Shock -
3 power was provided for the steam generator work via 480-volt power supplies. Statements from Blectrical Maintenance employees indicate that the chance of serious injury from this incident was remote. Electrical shock could have occurred if an employee had been inside the bowl of the S/G when the CENESIS manipulators " shorted out".
But this potential would have been overshadowed by the potential from the water surge.
0120R
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- E 22)-At the time of.the testing of 1-FCV-63-1 valve and subsequent release of water through the steam. generators, TVA had ongoing.
p
. work involving two contractors, Combustion Engineering and b
Westinghouse... Combust..lon had three GENESIS manipulators, one each in S/0 #2 HL, S/o #3.HL,- and. S/G 84 HL.
All manipulators were
' checked at 1600 EST on February 1, 1987 and found to be complete and properly positioned.: All manipulators were later checked by combustion Engineering personnel.and eventually removed for maintenance.
?. knob and a hook had care off the S/0 #2 HL GENESIS but we're' verified as remaining-in the channel' head af ter'the water spill. Subsequent maintenance has verified that an accounting has been made for all parts.
Westinghouse had two heat treat end effectors, one each in S/G #2 and S/G #3.
Each was locked onto*the GENESIS manipulator and;the one.in S/c #3 was cam-locked.into'the tube sheet.
In addition, a heat cycle was in progress on R2C57 tube. The heater and connecting conduit was removed at 0521 EST on February 2,1987.
All, Westinghouse items were ace 6unted for in S/c #3.
~
During.an earlier attempt to remove a, heater from RIC36 tube of S/G #2, the heater disassembled a'nd.two small pieces of Nextel insulation (approximately 1/2" long).were dropped into the HL
+
channel head, one other small white object was observed but the-camera could not focus well enough to positively identify the object as a piece of Nextel insulation ~or as a ceramic bead. All-objects were observed and video taped by use of the pan and tilt tubesheet camera. Because of the size, weight, direction of water flow, and the cha.nnel. head geometry,-it is probable that any 3_
debris was either washed through the manway or remained in^the channel head; however, there is the possibility that small debris O'
could have entered the'RCS through the (.64" diameter) channel head drain hole. As a part of the Westinghouse procedure, the stuck heater will be removed and examined for missing par.ts. A o,
subsequent inspection of the S/C #2 HL channel head after the e-water spill showed items of similar size and quantity still remaining in the channel head along with the knob and hook from P
the GENESIS..After the heat treatment was complete and the CL tube plug was removed, Westinghouse did successfully remove the stuck heater from Row I, Column 36.
During the removal process, a catch bucket was positioned under the tube opening and the stuck
~~,
heater was blown through from the opposite leg. This removal method ensured that all remaining parts were captured and that no additional loose parts were introduced into the channel head.
An inventory of the heater, including parts from the channel head, was performed and six ceramic' beads were identified as missing.
The amount of Nextel insulation missing was r.ot quantified.
During the removal process, it is probable that the beads were disintegrated and became mixed with the other debris resulting from the disassembly of the heater and therefore, could not be readily identified. A worst case analysis assuming that the ceramic beads end an assumed amount of 12.44 grams of Nextel entered the RCS through the.64" diameter drain or through the nozzle when the cover was dislocated by the water flow, has been furnished by Westinghouse (Attachment 16).
0120R
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- 23) The caution order ldid not contain sufficient information about the.
abnormal condition the plant was in at the time of the event. The caution order required th.e Uo to contact:the 53 prior to opening i
valve.
It did not requi'e 'any~ notification to the control point r
for'the s/o work.
Tagging procedures in MI-1.2 Appendix B,-M1-3.1 Appendix B, and the Mode 5 LOC'. requirements contained within AOI-6 and soI ~14.1A have been reviewed and determined that a Hold Order cannot be placed on the valves FCV-63-1,62-135 and -136.
The mis appropriately call for a caution order on the subject valves. The caution order has been strengthened as called for in the interim recommendations (Reference Attachment 11) for MI-1.2.
Additionally, as called for in th interim rect:.e ndations, the e.
breaker is opened on the subject valves whenever someone enters a so bowl. We evaluated the possibility of changing the clearance to call for a hold order but decided that it was not in the best~
interest of overall safety. Th~e operator must have inenediate availability of the'RWST via FCV-63-1 to mitigate Mode 5 events.
that may occur (such as the RHR spray event, the loss of RHR pump suction events that SQN has experienced or a Mode 5 LOCA).
~
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____-_____________1______
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VII.
RECOMMENDATIONS January 28, 1987 Event j
A.
l
- 1. A diverse level indicating system shall be installed on RCS hot leg loop R4 on the RHR 14-inch letdown line. A split range pressure transmitter system, with readout in the main control room, should be installed in parallel with LT-68-66 near the floor elevation (680 feet). The output should be calibrated in feet above sea level. The. narrow range - 693 to 697 feet and the wide ranga - 693 to 730 feet. A low level alarm should annunciate at approximately 695'-2".
- 2. An RHR flow rate correction factor for the above' level indicating system shall be developed and placed in applicable
~
~
procedures.'
- 3. The elevations of all pertinent features in the open reactor system (RHR nozzle, loop cen,terline, S/G manways, vessel
~
flange, purge vents, etc.) shall be determined and documented in an appropriate procedure. The as constructed field elevations should be obtained'on each S/G manway (refer to ) and incorporated into the appropriate procedure.
SI-693 should contain an acceptance criteria.
~
- 4. Enlargement of the existing 3/8" sense line to the sight glass shall be evaluated following the results of the periodic flush and comparison check with the tygon tube.
- 5. Interviews wit'h various plant personnel revealed that a' Black Beauty (sand blast material) type substance was removed from the site glass line; Analysis shall be conducted on any O'
substances removed from this line as a result of the weekly flushing operations.
- 6. RCS level shall be maintained between 695'6" and 696'1" while the SG manways are removed and nozzle dams are not installed.
The loop centerline is elevation 695'.
The top of the inside of the hot' leg pipe is 696* 2 1/2" and the cold leg pipe is 696' 1 3/4".
Any time the water level is above the 696' 5 1/2" mark, it becomes a lot more sensitive to increase in fluid inventory since the loop piping is full of water.
B.
February 1, 1987 Event
- 1. Clear lines of communicate ~ons shall be prescribed by applicable procedures and established between control room operators and individuals opening or working on an open RCS.
This communication may be effected by controllers located at a designated control station that in turn are in communication with the individuals at the work station. This includes the l
process of opening the RC9 and subsequent work activities in the area of open S/G manways, open reactor vessel, removed pressurizer safety valves, etc.
0120R aac 4
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- 2. The communication link shall be utilized to notify the centrol room operaters of the start / progress / completion of the subject l
work activities and the control room operators shall notify the local work stations' of. all system's perturbations., whether
" caused by operator action,-equipment failure, transients or accidents, and prescribe actions to be taken (evacuate, etc.).
The PA system shall also be utilized by operations to notify pla,nt personnel of such events.
- 3. The stroking of electrical and air-operated valves in the SI-166 series chall receive a review by a licensed operator.
Detailed valve alignments, operating conditions, and j
precautions should be specified for testing, if appropriate.
A two week limit should be placed on the duration of a Masterlist to ensure the data is reviewed by the Mechanical Test Section in a timely manner.
- 4. The upcoming training on the.SI program shall utilize this
~
e' vent as an example where an inadequate instruction existed and the consequences of not stopping to correct the instruction.
~
- 5. Specific training shall be conducted to inform operators of the sensitivity of the RCS level during the time the S/G manways are open and the potential for flow by gravity from the vcT or RVST.
- 6. The manager / engineer responsible for the steam generator work shall review t.he clearance with the ASE and UO.
Speciai emphasis should be placed on pumps, valves, etc., not tagged with a hold tag: All caution orders shall be reviewed and concurred with by both operations and the supervisor 54 responsible for the job.
~
- 7. Several problems were identified after the incident in the overall set-up of the steam generator work. To alleviate l
i problems of this type:
l I
all electrical cords should be routed overhead whenever
{
a.
possible.
m b.
breakers, valves, and other disconnects for electrical and pneumatic power for steam generator work should be legibly marked as to what the disconnect / valve controls.
0120R aac musummuisuum u i i
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- 8. The information in the current night orders should be included
[
in a plant procedure to be followed each time the steam generators are opened.,
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9.Operationsemployeesshal1 reinstructed' tor $viewflowprints when performing valve alignments.
- 10. A detailed i.tview shall be conducted of the procedures and clearances for the secondary side of the steam generators prior to the next wrsrk involving breaching this system.
- 11. A procedural'centrol system shall be established to ensure important information is not deleted from a procedure during the-revision process.
C.
peripheral Issu'es
- 1. Identify all lighting cabinet" breakers, especially those inside containment, and label as appropriate.
~
- 2. Excessive. thrust can be applied to motor operators and valves' when closed manually. This can cause damage to the valve and
~ The excessive thrust could possibly keep operator internals.
the operator from reopening the valve if needed. Under f'
non-emergency conditions, a valve should be declared inoperable until it is manually reopened and electrically stroked.
This will ensure that no operator damage has occurred when manually closing a valv,e.
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The most significant part of this event is that the RHR pump 1A-A had to be stoppe.. A followup interview with the DO on February i
27, 1987 determined that,the RHR pump miniflow valve was starting to open as the operator Was~ securing"the pump. This meant the pump flow rate had decreased to approximately 500 gpm. Based upon the discovery a second pro (1-87-095) was initiated as this event and'a four hour notification was made in decordance with 10 CFR 50.72.a.2.111.
LER (50-327/57012) was submitted on the same day in accordance with 10 CPR 50.70.a.2.v.
These notifications conservatively assumed that both RHR pumps were inoperable due to the low RCS water level. The loss of decay heat removal is the subject of numerous experience review documents (references 25, 26, 27, 28, 29, 30). While the weekly flushing and daily tygon tubing checks will increase the tsliability of the sightglass -
level system, an independent and diverse level indicating system is a must for the adequate long-term resolution of this problem.
Additionally, our investigation concluded that the S/G lower head will be wetted at elevation 696'8" and will spill over at an estimated 696'10"-(refer to Attachment 5).
The hot and cold leg.
loop piping are full of water at' elevation 696' 2 1/2" and 696' 1 3/4". respectively. Therefore, the upper limit for water level is recommended as 696' 1" to help decrease the sensitivity of the water level to fluid inventory changes.
The root cause of this event is therefore determined to be lack of a diverse level indicating system providing redundant information to the operator. The open reactor vessel mode of operation was o..
not considered in detail in the plant design. Based upon our investigation, we feel that the recommendations in this report, for this event", provide adequate margin to prevent recurrence of this type of event.
o.
B.
February 1, 1987 Event 0
Based upon our investigation, we concluded that the operator was P
stroke testing 1-PCV-63-1 at the direction of Mechanical Test Section for increased frequency requirements of refe'ences 1 and 2 r
when actually, testing of the valve was not required. The valve had been regeared to slow it down. The operator realized the instructions given in SI-166.3 were inadequate, however, the operator attempted to supplement the instructions by stopping the RHR pump and closing the pump discharge valves without closing the suction valves from the RCS t,o the RRR pump. This provided a flow path from the RVST into the RCS.
The operator opened 1-PCV-63-1 and water flowed by gravity into RCS hot leg loop 4 until the valve reached its full open position and was immediately closed.
The valve was open a total of approximately 42 seconds. An estimated 6000 gallons of water flowed into the RCS and approximately 3000 gallons spilled from S/G manways.
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one individual was on the s/G s2 platform when the water started flowing from the mt.tways. A total of 6 people were inside the polar crane wall. Although the potential for serious personnel injury was possible.' everyone reacted in a professional manner and injury was avoided. HP responded properly to the event by relocating the step-off pads, evacuating people from inside the polar crane wall, and adjusting RVP requirements. The containment air monitors (1-RM-90-106/112) showed no appreciable increase in activity. Air samples later confirmed there was no increase in activity due to the spill. Exposure for the cleanup resulted in 4.82 man-rem exposurs. While it was suspected that 1-PCV-63-1 did not fully reclose, MOVATS testing and leak rate testing confirmed it had fully reclosed.
~
Wo reactor cool' ant system dilution occurred during this event.
There was no damage to plant equipment and all the S/G equipment -
was subsequently repaired and returned to service.
The" investigation concluded that 1) an inadequate instruction,,
- 2) not stopping work and obtaining an. approved instruction change,,
and 3) not referring to flow dia' grams prior to valve manipulation caused the event. The inadequate instruction was determined to be caused by mist'akes in the revision process that allowed the operating conditions for testing to be revised and later deleted.
The root cause is therefore determined'to be the lack of a procedural control system to ensure important information is not deleted from a procedure during the revision process.
Additionally, the, comniunication between the Main control Room and the people working on the S/cs was not adequate. This inadequacy was resolved prior'to restart of S/G work (refer to Reference 24).
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IX.
REFERENCES 1.51-166.3 - Full Stroking of Category "A" and "B" Valves During Cold shutdown'
-. ~
2.
SI-166.8 - Increased Frequency Testing of Category "A" and "B" Valves 3.
SI-673
- Reactor Coolant System Level Verification Using Sight Class of Tygon Hose 4.
OSLA-58
- Maintaining Cognizance of Operational Status 5.
Interviews conducted with:
Main Control Room Operators. (00), (ASE),and(Sk)on2/2/87 a.
concerning the 2/1/87 event.
'Two AUOs on 2/2/87 concerning the 2/1/87 event, b.
Three HP Technicians and two laborers on 2/2/87 concerning th'e c.
2/1/87 event.
~
d.
Three Westinghouse individuals on 2/2/87 concerning the 2/1/87 event.
S/C PM on 2/2/87 concerning the 2/1/87 event, e.
TVA S/O Suppo'rt Engineer on 2/3/87 concerning the 2/1/87 event.
f.
W Electricai~ Maintenance Engineer on 2/5/87 concerning the g.
MOVATS testing of 1-FCV-63-1.
h.
U0, ASE, and SE on 2/6/87 concerning the 1/28/87 event.
i.
Three HP technicians on 2/6/87 concerning the 1/28/87 event.
3 j.
Maintenance Planning Supervisor on 2/6/87 concerning the SI-166.6 Masterlist.
k.
Operations Manager on 2/6/87 concerning the SI-166.6 Masterlist.
6.
Meeting in Plant Manager's office on 2/5/87.
Meeting in Site Director's office on 2/5/87 and 2/6/87 7.
NRC Inspection Report 50-327-50-328/86-59 and TVA response to item 86-57-01 (L44870123 805).
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8.
Flow Diagram 47W811-1.
9.
Flow Diagram 47W812-1.
=
- 10. PRO 1-87-046
!!. SOA-lC6 completed 1/29/87.
- 12. Daily Journal of the unit 1 BCT Control Foint for the following dates:
a.
January 28. 1987 b.
February 1, 1987-
- 13. SQA-186 completed 2/1/87.
- 14. PRO l-87-056.
- 15. Daily Journal of the Unit 1 po/ASE (SRO) dated '-
1/21/87 1/27/87 1/28/87 1/29/87 1/30/87 1/31/87 2/1/87
- 16. Daily Journal of the Shift Engineer for 1/27/87 1/28/87 2'
1/29/87 1/30/87 1/31/87 2/1/87
~
- 17. Night Orders issued by the Operations Manager dated 1/9/86 (SIC) really 1/9/87 2/1/87 2/6/87 2
- 18. SI-166.3 - requirements for stroking of FCV-63-1.
- 19. Informal memo from W. D. Romine to J.11. Sullivan "SRP - MOVATS Testing of 1-PCV-63-1" dated February 7. 1987 (S53 870206 841).
- 20. SI-673 Data Sheet I for 1/27/87 and 1/28/87.
- 21. Caution orders 1082 and 1083 for 1-FCV-63-1 and 1-PCV-62-135/136 respectively.
- 22. Hold Order No. 1084 for Unit 1 RCS.
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- 23. Informal memo from David F. Ooetcheus to L. M. Nobles dated
)
February 6, 1987, "SWP - Implementation of Water Spills Task Force
(
Interim Recommendations" (500 870207 801).
- 24. WRC Case Study Report AB00/C603, Decay Heat Removal Problems at
-- U.S. Pressurized Water Reactors" December 1985.
- 25. NRC Information Notice 81-09, ' Degradation of RitR System" dated March 26. 1981.
- 28. INPO SER 86-35, " Extended less of Shutdcun Cooling Due to Steam Binding of Shutdown Cooling Pumps".
- 29. INPO SER 79-84, " Loss of Shutdown Cooling Due to' Inaccurate Level Indication".
- 30. PRO 87-095
- 31. LER 327/87012
- 32. LER 327/87013
- 33. SOI - 74.lA R38
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- 34. AOI - 6 R18 7
- 35. MI - 3.1 and 2.1 W
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X.
ATTACHMIDiTS 1.
Sketch of S/0 Laydown Area 2.
Sketch of S/0 Layout 3.
S/0 Manway Shield Door 4.
S/0 Manway Shield Door - Filter Connection Side S.
Diagrams of S/O Lower Head (Manways. Nozzles) 6.
Staplified Flow Diagrams of February 1,1987 Plow Path 7.
SQA 186 completed January 29, 1987.
8.
SOA-186 completed February 1, 1987.
9.
Night Orders issued by the Operations Manager dated 1/9/86 (sic) really 1/9/87 2/1/87 2/6/87
- 10. SI-166.3 Revisions for changing requirements for stroking FCV-63-1.
Task Force Interlm Recommendations" dated February 5, 1987.
- 12. H.sndouts for the 2/5/87 and 2/6/87 INFORMATION SESSIONS.
Sh
- 13. Training Attendance Roster for the 2/5/87 INFORMATION SESSION.
- 14. Training Attendance Roster for the 2/6/87 INFORMATION SESSION.
~
- 15. SI-166.6 Test Data Package for Testing 1-FCV-63-1 on 2/4/87 due to e
WR B221062 and ICH 87-117.
- 16. Westinghouse response to Non-Conformance Report TVA-87-00023, Unaccounted Heater Beads with safety analysis attached.
- 17. Recorder Chart 1-TR-74-14 for 1/27/87 through 1/28/87.
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g The Hoot Cause Assessment f4CA) included in this procedure in designed j.
to be useable for the investigation of all plant events, pons i'j Assessment Section w!!! avaluate the RCA for concurrence. Thin' l
prnecoure le f or root cause identification only - consequene:en of g
advesse condition must be evalasted per SQA84 (PROS) and SQAll8 (CAQs) 1 for flRC deportability and safety significance.
J I
2.0 PUHPOSE g
I*
2.1 The purpose of the RCA is to aid in the determination and documentation of significant root cause events. The analysis will II l
provide a standardized format so that consistant techniques and i
terminology will be used in RCA.
2.2 The following is a listing of possible events which should be
}.}q investigated for Root Cause s however, the plant staff can use 1
this procedure at their discretion, for other types of i
p l' events /condtlons.
h',
Sigificnat plant off normal conditions a.
t..,
b.
Hajor equipment f ailures f*
c.
Generic equipment failures j
d.
Design deficiencies.
3.0 REFEki:llCES o f..
.as 3.1 IHPO 84-027 "An analysis of root causes in 1903 and 1984
'9, significant event reports".
3 j,
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3.2 Commitment Action Tracking (CAT) system item no.85-475.
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der! HIT 10tlS N, ' ' },*
4 3. '.
Adverse Actionst Personnel Actions that initiate the event or lead it F.].
In t;he adverse direction.
Adverse Condition: A condition that ititlates the event or leads it in 1
the adverse direction.
3,
i
' f. * ;
Root _Caunt The most basic, fundamental cause that can be identified
+
/j.d.,aI for the adverse actions or conditions, using the information that is
!h) obtainable during an event investigation.
I.f"j 5.0 perform the rnot cause analysis in accordance with Attacivaent A.
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1.0 17D15 TilAT Sil0ULD DE Cor SIDERED DUR!tlG ROOT CAUSE ASSESSitDIT s
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'a.
Equipment fallure
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Personnel error
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Procedural errors I ',,' ' ', '
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Inadequate training t
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Design errors e
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Abnormal environment
'e 2.0 PRrnrQUISITE
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Obtain a thorough understanding of Attaciuner t D.
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' 3. 0 2DDIT!!*1 CAT!0rl DATA
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.3.1 Provide the identification number for the applicable plant J. 8 l l.,i
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activity (l.e., PRO / Instruction / Report,tio.)
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3.2 Record the unit (s) and system (s) for ><hich the adverse a
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action / condition is applicable.
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3.3 Record the start date for the assessment.
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3.4 Reference material used for assessment (i.e. technical bulletins,
[8' - li. '; r vendor manuals,
.etc.)
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4.0 Dr.5CRIPTION OF EVDTI'
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l' Provide a brief descriptten of the event that is being essessed or M. i reference a record which.streedy containe e description of the event.
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5.0 ADVERSE ACTIOilS/C0ftDITIOflS AND ROOT CAUSE
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, f, Driefly dancribe the specific adverse actions or conditions that will 4
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5.0. ADVERSC ACTIqtl}/CONDITICilS AND ROOT CAUSi: (continued)
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- significant - events a\\
re of ten too large o'r complex to allow ^tnenedi tate j
!.1 identification of the roch cause.
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Therefore, the root cause. process 8* k involves identifying the Jodivldual adverse actions or conditions for each -
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.Some events involve snore *than one adverse action or condition.'
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. ( ? described below.'
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n r.ach adverse action or,' condition may'be completely independent of the 4
. M ir' '. !
1 others. In.that case, independent 1 -
IMf,4 7,:t ".,
identitled, for emernple, a sensor Individual root causes enould be g - D, *,,C might be found to be both connected wrong (root causes ; failure to follow procedure) and sainen11brated. troot i y, g.; lj j '
causes do!!=lont, erroneous procedure); either. item could have caused j
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.the event.'
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- An' adverse action er condi'tlon may be dependent on another adverse'
-j i
ation or conditlons neither action or condition by itself would have
, !,! !.6.lf,f;'.' J.$.
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caused the event, but botti were needed.
. I. 4
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should be-Identitled for each item. ror'exampid; maintenance enight bei In that case, the root cause-(
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, performed innproperly (root causes lack of training) and the problem i,.,i. '!i j ;
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might not be discovered due to Inadequate post-maintenance testing
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(root causet deficient, Incornplete procedure).
p' t,. )h. '.
Thenri are considered to be. dependent because the probiern must both occur ~ and be undetected.
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The adverse actions or conditions that are identified may all be found I
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to stem from the same root cause.
i i,. ' L Tconditions should be combined into a single, higher-level adverse-In thls cas
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action or condition. For exarnpt y 1,'. hj il,.,;p ' ' ' identitled might include failurg,e, the adverse conditions that are q
to open, and failure of certain' Indicators, all cauned by tho' loss of aof a M
l l.
j g. *e eingle power source.
- Bather than identifying the same root cause for MP/.
ir,( i f,g,l ',{ } I,p,"8 i each condition, the adverse condition should be rewritten as " loss of power source caused failure of pu.opi valve operator, and
., y.,, g *
of power. instrumentation," and the root cause should be identified for the los s
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barriers that failed and a!! owed the event to happen.The adverse
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condition on the event (e.g., technician misadjusted a limit s
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during calibration).
. root cause of the adverse action or condition (e.g., erroneousThe cau
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-3 adverse actions o'r cos:ditions that are Identified should be directly
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The
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related'to the significant aspects of the event.
Problema not related to
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the significant aspects of the event should not.normally be identified as
'8 adverse actione.or. condition.1.,.. I or example, if the significant event were
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.phat a safety system did not actuate following a reactor trip, then the
[ pl..j *,' ! * '.cause of the trip (e.g., IGC.,t,echnician error) would not be listed as an 6
adverse action. Ti.e adverse action (s) would be related to why the safety
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y system failed.to actuate (e.g., fuses were not installed fo!!owing
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The root cause-Is the roost basic, fundgrgptal cause that can be identified
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I for tfid'adve r'se'!sdtlon or.conditio'n; ~using 4he inforwiation'that is obtained 1 j'
., @ s[;.b during an event investigation. The root cause analysis effort should be taken to the anost detalled level possible. The following are examples of gq;{[.4 g
root causes j
t,,,
f:l;,
(1) A ulck analysis of an' event may l'dentify the. root, cause as f ailure to
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follow procedures,however,,,a, detailed, analysis might' find the root,
. g., ll V7..': 3 to.be fatigue due to excessive overtimo.
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(2) A pump falls during operation. A quick analysis.of the event could t.
et conclude that the pump failuu ;,.tcauso u. a worn bearing.
A detailed j*e g.G.' l;. ;
analysis of' the event could reveal: that the bearing Talled because the
/
34 91i.
wrong type lubricant was used. This could have a numbor of-root causes l l M IiEl {'.,-
(ie, inproper procedure, inadequately trained personnel or purchasing 8'
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failed to buy correct lubricant.
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The Root Cause Assessment (RCA) included in this procedure is designed to be useable for the investigation of all plant events. PORS Assessment Section will evaluate the RCA for concurrence. This _.
procedure is fbr root c5 usa identification only - consequences of.
adverse condition must be evaluat.ed per SQA84 (Pros) and SQA118 (CAQs) for NRC deportability and safety significance.
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2.0 PURPOSE 2.1 The purpose of the RCA is to aid in the determination and - -
documentation of significant root cause events. The analysis will provide a standardized. fomat so that consistant techniques and termino' logy will be used in RCA.
2.2 The following is a listing of possible events which should be investigated for Hoot Cause; however, the plant staff can use this procedure at their discretion, for other types of ovents/condtions.
a.
Sigificnat plant off normal coaditions b.
Major equipment failures
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c.
Generic equipment failures e
d.
Design deficiencies 3.0 REFERDICES 3.1
's' INPO 84-027 "An anilysis.of' root causes in 1983 and 1984 significant event reports".
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3.2 Commitment Action Tracking (CAT) system item no.85-475.-
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4.0 der!NITIONS
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m Adverse Actions: Personnel Actions that initiate the event or lead it C' ' %
in the adverse direction.
O Adverse Condition: A condition that ititiates the event or leads it in
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the adverse direction.
s Root Cause: The most basic, fundamental cause that can be ideni.ified for the adverse actions or conditions, using the information that is obtainable during an event investigation.
5.0 Perform the root cause analysis in accordance wi.th Attachment A.
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Page 1 of 5 RCA EVALUATION FORM RCA Tracking No. -
(To be assigned later by POMS)
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1.0 ITDiS THAT SitOULD DE CONSIDERED DURING ROOT CAUSE ASS 4
a.
DTulpment failure
.b.
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Personnel error.
c.
Procedural errors d.
Inadequate training
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Design errors -
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Abnormal environment 2.0 PREREQUISITE Obtain a thorough understanding of Attac
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7 3.0 IDDITIFICATION DATA m
3.1 activity (i.e., PRO / Instruction / Report No.) Provide the ident 3.
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d l~ - /6 6, 8 3.2 Record the u it(s) and system (s) action / condition is applicable.
for which the adverse.
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=:s Record the start date'for the assessment.
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It ah/a7 3.4 Reference material used for assessment (i.e vendor manuals.,etc.)
technical bulletins.
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4.0 DESCRIPTION
OF EVDtr Provide a brief descr8.pt2on of the event that is being assessed or reference a record which stready,contains a description of the event.
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5.0
,T ADVERSE ACTIONS /CC?IDITIONS AND ROOT CAUSE
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Briefly describe the specific $dverce actions or conditions that 0111 be analyzed and deterinine root cause.
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Page 3 of 5 5.0 ADVERSC ACTrotis/coNDITrons Arm Rcor CAUSE (continued) 5.2 Adverso ActionCendition #2 p [4
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Page 4 of 5 6.0 CONCt.USIONS A*ID RICOMMD!DATIONS: --
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.2 2 I o t, Z - Completed Dy: . MMOG. R e, o/f/,s7 Date Reviewed By: / f D) t -- Date d [ [ p 7.0 Assessment Section for evaluation.A copy of the Root Cause Analy' sis la to b r. s.,,. Transmitted to PORS d By: '~ Date: c /; e, I we l 0229s/taf he e es.IPS
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1 g:.. i. ' Qaecjg .~ SQtlP SQA186 Page 6 Revision 0 .] i ATTACMDIT' A - l ..r. .: s Page 5 of 5 12.0 PORS ASSESSMDrr SECTION EVALUATION /DH'OMMDIDATICNS* t u. e ~~ / PORS Representative Date
- The evaluation will include an asset:sment of the root cause for each adv'erse action /condtion and the conclusions and recommendations for concurrence.
13.0 Assign a permanent tracking number to the ccver page of the RCA, log s:t brief description of the RCA and its tracking No. in the RCA log book. Logged by ~ / 14.0 Place the RCA in a clearly labled folder and file. m 4 D. C' 02295/tdf 1 s Dk
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.\\,, SQNP l t, SQA186 J page 7 lj -) : Revision 0 -i Kt ta'chKenC"Ji ' ~. ~ Page 1 of 2 1 {i . identification of the root cause.Significant events are of ten too large or complex to + . Therefore, the tr'ot cause process involves identifying ~ the individual adverse actions or conditions for ea h
- event, Some events involve nore.,than one adverse action or condition.
c Possible relationships betweet: + described below. multiple adverse actions or conditions are Each adverse action or cond' tion may be completely independent of the' - i
- others, In that case,. Independent, individual root causes would be
' t. identified. For example, a sensor might be found to be both connected wrong (root cause: deficient, erroneou's~ procedure);.either item could have ca cause: the event'., f An adverse action or condition may be dependent on another adverse ation or condition; neither action or condition by itself would have caused the event, but both were needed. .; : g
- should be identified f6r each item.
In that case, the. root cause 'l performed improperly (root cause: For example, maintenance might be' lack of training) and the problem I might not be-discovered due to inadequate post-maintenance testing j troot cause deficient, incomplete procedure). be dependent because the problem must both occur and be undetectedThese are cons The adverse actions or coriditions that are identified may all be fou$d to stem from the same root cause. action or condition. conditions should be combined into a single, higher-lev For example, the adverse conditions that are Identified might include failure of a pump to start, failure of a val i to open, and failure .l
- single power source. of certain indicators, all caused by the loss of a ve each condition,.the adverse condition should be rewritten as " loss o s
power source caused failure bf pump, valve operator, and e instrumentation." and the root cause should be identified for th ,4..' of power. e loss 4 barriers that failed and allowed the event to happen.The adver ? e o condition on the event (e.g., technician misadjusted a lim The mechanism of r i during calibration). root cause of the adverse action or condition (e.g., erroneousThe cause f> ~ procedure). e 02293/tdf O
n, i.. 9'g l.' soup SQA186 .Page 8 L 6 Revision 0 C .-.:. =- ~ Attachment B - Page 2 of 2 s' l The adverse actions or conditions that are identified should be directly related to the significant aspects of the event. adverse actions or conditionsi,the significant. aspects of the event sho that a safety systera did not actuate following a reactor trip, then' theFor s cause of the trip (e.g., 'I&C., technician error) would not be listed as an adverse action. system failed to actuate (e.g., fuses were not installed following ._ maintenance). The root cause is the most basic, fundamental cause that c e during an event investigation. taken to the most detailed level possible.The root cause analysis effort should be root cause: The following are examples of (1) ~ A quick analysis of an event may identify The root cause as failure t ".. ) follow procedure; however, a detailed analysis might find the rooto l cause to be fatigue due to excessive overtimo. '2) A pump fails during operation. conclude that the pump failed because of a worn bearing.A quick analysis o. t analysis of the event could reveal that the bearing failed because the A detailed wrong type lubricant was used. tie, inproper procedure,. inadequately trained personnel or purc failed to buy correct lubrica nt. 2. , I. SY Gg e 4 l 4 o 0229S/tdf 1 f mes W g ge c=
'9 .h# ~ hNo.(, lot 8,nlh ytteth9 us SEQUOYAll NtlCLEAR PLANT ~ OPER ATIONS GROUP $ pl, Page 3 b OSLA30 06/11/86* (- ATTACllMENT 1 h4cP l OE h y INFORMATION OR TEMPOltAltY INSTlWCTIONS FOR Tile OPER,ATIONS GROUP The attachedLt.ER_is potentially cligibic for a $50,000 fin,c and a re occurrence of this cyp _? of event could result in escalated enforcement actions. Mechanical Test expects to have revised procedures approved short l y, but in the interim, note the requirements and the specific valves ~ requiring testing immediately af ter hold order release. = ~~ ^ //,9!f6 M._ w- / Signa t u re gate Distribu tion : 2 SillFT ENGINEElf All Alfo STATION M TitAINING SE ~ M. U-l MCR ~~ Til AtrO STATION Ngw NAgglip gj M U-2 MCR
- cowp, Til ASE COND DI AtlO STATION X
POTC - PWR UNIT Sl;P\\' SillFT ENGINEER'S OFFICE rh 4.4. e
~ ^ January 7. 1987 I
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~ - Jarvis Anthony ~
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POB-2 ASME SECTION XI STROKE TIME val.VE. TESTS We have had both recent and rei~urring problems with our ASME Section XI Stroke Time Testing ptogram. A recent LER and notice of violation are attached which are indicative of these problems. We realize our procedures are confusing as they present-ly exist and a major revision is in progress to clarify the requirements. To inforn you of short term potential problem areas requiring attention prict to our new procedures being issued...the_.folloulag information is provided. Each valvc' required operabic must be~tes,ted at the required, frequency
- to be considered operabic (generally every 92 days).
In order to meet this requirement all valves required to' be operable shall be tested within the 25% extension time of the packarc. A list.of valves required operabic in*c. ode 5 was provided to you carlier anel another copy is attached for your information. Valve 2-FCV-62-138B is presently tagged out and can't be operated. This valve is required operabic in mode 5 and was, last tested 9-26-86 Thin valve must be tested ac soon as t h,: 1.old order is removed to be declar'd operabic. e Valve 2-FCV-74-24 is also prcsontly tagged out and was last tested 9-25-86. This valve must be tested as soon as the hold order is removed since it is required operabic in mode 5. SI-166.1 for Unit I will be issued 1-14-87; SI-166.3 for Unit 2 will be issued 1-19-E7; and SI-166.21 for Unit 0 will be issued on 1-20-87. These. packages contain node 5 operability valves to be tested within 25% of the issue date to avoid another LER. Please issue a night order addressing thoce items to p r o.' i d e interim administrative controls until our new procedures are issued. Contact David Barker at 6634 if any questions arise. & kJ H-& U. H. ':a c kay ~ ~] WHM:KRL Attachments: 3 1 e 9mMup
/, TENNESSEE' VALLEY AUTHORITY as. gva asm eru ots.t.rs sweet or SeQuNfN Ytt lWC e$ oh ( eus,eer L - e n o,n c, n e_ J CowPWft0 to catt CH4C at o 87
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- i Ol'Hil ATIONS GitOUP Page 3 OSt.A30 00/11/85
~ ATTACllNENT I ~ -INFORMATION OR TEMPOlt AltY INSTitUpTf 0NS-itOR TilR OPERATIONS GROUP j
Subject:
Unit 1 Reactor Coolant System Level During the on going m\\odif;:ation to the Unit i steam genera tors," it is f.- the responsibility of the Or,erations Section to maintain the RCS level in a stabic condition between Elev. 695' 6" and 697' to prevent over-flowing thru manway cover openings on the primary side, There have been two events in the last week where water has overflowed out thru manway covers which could have caused injuries and contaminated personnel working in the area. Honetary expense has been high due to damage to equipment and delays in modification work. In the future when making changes to any equipment that might affect the level in the RCS, review and approval by both the U-l ASE and Shift Engineer is required. Prior,to the planned operation notify the Combuction i Engineering Trailer at PAX 7102 or 7238 so their people can cicar the steam generator platform. It is best not to make any changes to the RCS, until work is complete in 'y l' the steam generators. J f/,,,, * \\ \\ ._ _ sk. kY& _ '~ / 5 597 Signature (, ~ Date Dist ribu tion : 3 Slf!PT ENGINEER R All Allo STATION I P TRAINING SE b-2 U-lMCR Til AUO STATION [ NEW M A KI'.t'l' DI U-2 MCit CDWR Til ASE ~~ COND DI AUO STATION l'OTC - PWR UNIT SUPV ~~ SillFT ENGINEEll'S OFFICE m ed N . as a. e .e .* ] l \\ S e g 6
a .Oc..cf 6.~ % ' v.. yt.,r.;.. p1(t[..;$.!)(!l o ..1... 'SEQUOY All 'ilVCLEAfl l'LAllT ,- y r ' '; OPEllATIOlls Gl100P ' fp,..., P w,. 3 ) qq OSI, A.10 .'\\ Y, p . h *,' ' " '. l/ l 06/11/05 ,.., ATTACllBIEllT 1 i< gl. i _ -. _: __. __ L. g lhf.'. 'lllFORfIATIOff on TEblPORARY lilSTRUCitoils FO'R Tile OPEllA'llOffG C:tOU ~ f- .. h,, .* Sulaject, VillT 1 STEAll CEllEIMTOR ACTIVITIES. ,{ 8 .ud..... : h.' ' See atenched. ,,l '. '.,t
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, Dis t rib u tion :, n'.- ,;;v y'j. <..'"': T SiiiTT EliGitiEI.7 AB Allo STATIOll G TD'AUO STAT! Oil 'I-TR AlillilG SE r p l"/, y - U.! IICI! [ llEW fIAKEUP DI ~~ . e, U - 2 l I C.'. CDWE ~~ TD AST. 1 POTC - PWit UillT Gill"/. COllD DI AUO STATIOll SillFT EilGlilEER'S u TIC - 1.
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i Subjects Utl1T '1 STEAll CEHERATOR ACTIVITIES - = - a-
- 'n action to prevent recurrence of the recent over flowing of HUS thronnh plenm i
generator menways must be estabtiehed. Clear linne o f c oeneaues t e n t t on munt tec was h ing use or around es tablished between cont rol room operatore 'and 'Individun ts [ This communication may be effected by the open steam generator manwara. , j 'i g a designnt.ed control station (C.h. trailer) that in turn . controllers located at I are in communicat ion with 't he inoividunin at the work station. 'the conminicat ion I ! link must be utilized to notify the control room o'perators of the start /pronresn/ I* 'lcompletionofthe ntenm generator work activitics. This will be accomptinhed jeachshiftanddocumented, ~ (C.E. c on t r'o l ~ ~ ~ i , The control room operntors must notify the local work stations trailer at extension 7102 or 7238, elevation 690 control point at ent'ennion 6349) l I
- of all planned system level changes and system perturbation. whether couned i by operator action.. equipment' failure, transients, or accidents, and prencribe action to be enken (evncuate from inside the criinn wait).
The pA nyneem shall also be ut111 cd by operators to notify plant personnel of nucli events. The unit operator or ASE is to log all communicat ions made wi t h t he ntram _y.. generator controllers concerning their activities and notification made. I, 3*- Additionally, to ensure the steam generators will not.over flow wlun nn coordination between individual is going to work inside the steam generntors, ~ for 1 - rcy - s.1 - 1, the control room and C.E. controllers w111 require ACDs l-FCV-62-l M, and 1-FCV-62-136, to be opened prior to anyone enterine. the steam generator primary. side with an A'SE re'maining in proximity of the breaker I in communication with the main control' room. When controlicrs compartments $), fnotifytheunit operntors that the individual hna exited the steam generator. for above mentioned vnives. l
- - l g the ASE can reclose the ACDs i
, When making changes to any equipment that might affect the level in the.RCS, review and approval by both the unit 1 A$E and Shif,t Engineer is rerpelred. Documentation can be by entry intp the 16g for approval. 4;
- At least once per day, on operator should valve in the tygon tubing.for vessel level and compare Its icvel with the Icyc1 gauge in the fan room.
Comparison f. data is to be logged in unit 1 unit operator log. If ponnible, this comparison should be done cach shift. As a mlntmum, the permanent gange glans in to bc isolated, drained, and refilled once per week on Sunday 3-11 shift. Documenta-tion of this event will be by entry into unit I unit operator log. l a. I g e.., s N '.. )., .m ..,. 7.. 3,. -+'--- -
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7.9.1 Operaties conditions for testing: I The plant is la cold shutdown and RCS pressure is i 1 30 pais. { l
7.9.2 Precautloos
7.9.2.1 Valve TCV-63-1 isolates both Irsins of RER' f rom the. RierT. i 7.9.2.2 Valve FCV-43-5 isolates both trains of SI fron the RVST.
7.9.3 Jostructions
Complete Data Sheet 63 8 in conjunctive with the following I 1 Turn E3-63,(1) to open. ~ 7.9.3.1 l 7.9.3.2 Verify the red light ce and the green light E off. 7.9.3.3 Turn RS*63-Le close. Measure the stroke g Em. 7.9.3.6 Verify the ar-. I'. . -..i ' '. olt. II) Rafer to the reference sheet for the handswitch number } 5 5 4 g 4 j ..J ~
,g'%%/....#..., '?r ti,f r,. ,, e . ~..... ~ u.,.. _*.. ke 2.of 6, r. t.i 1 l--- m,vi.i 1 7.9 Oroup 63-4 Valve Numbers: 1-FCV-63-1, 3 m , 7.9.1 r4'erating ceeditiosa for tvettags I The plant is,in cold shutdown sad RCs pressere is 1 30 9s1. 3 7.9.2 Precautions I Valve FCV-63-1 f oolates both tritu"of RER 7.9.2.1 from the RVST. 1.i 7.9.'7.*2 Valve 'TCV-63-5 isolates botA trains of 31 - from the avs7. f
7.9.3 Instructions
~ ~ l Complet* Data Sheet 63-8 in conjunction with the following 7.9.3.1 Turn Es-63 [ to open, ~ ' 7.9,3.2 Verify.the red light en and the green light off. ~ 7.9.3.3 Turn ES to close. Measure the stro b ' B 7.9.3.4 Verify the green light.es sad the red light . ].- off. E e (1) Refer to the reference sheet for the bandswitch number f1 i 6 \\ l ~ Q ,o a v..
n;.. ij,. 'yg a "fl, -i.y '_ ' 1 l wy, w,.v. -r g. .,i l 1 p 7,g t* I O /. 'l ' ' $1 166.3 - Units 1 & 2 ._ Page 1 et 1 ~ '" Revision 5. /) 3g p{ l ~ 7.9 croup 63-8 Valve Numbersi ICV-63-1,5 s -- l 7.9.1 0;,erating conditions for testing: ) i The plant is to cold shutdown and RHR is isolated l l.
7.9.2 Precautions
from RCS. f 7.9.2.1 Valve TCV-63-1 isolates both trains' ofRXR
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- from the RVST.
V*. rm!, 7.9.2.2 Valve TCV-63-5. isolates both trales of SI l 1 from the RVST. ~ i ,,3 -
7.9.3 instructions
e. e. L t-a: L. .( complete: Data Shee't 63 8 To Voofunction with..the i following: _ lI - Tur,n, ES-63.(1) 1 .to*6 pen.. [ l 7.9.3.1 7.9.3.2 Verify the red Ifght on mod the green Ifght l off. g 7.9'.3.3 Turn RS (1) to close. Heasure the stroke [ ~'~ time. 7.9.3.4 Verify the green light on and the red, light off. ~ .ie t e t to the reference sheet for tbc bandswitch number 3 ~ .+ D ~ 20 .,. 5 ..n w ;
i 000339-3 I I. 64.3 - Units 1 & 2 Page 1 of 1 Aavistoa 'f,,, __, ' Cs O O ~ n,. ~ ~ 1 7.9 Croup 63 8 Valve Number's: FCV-63-1, 5. 3 7.9 1 Operating conditions for testing: The pisot'is in cold shutdown g. i
7.9.2 Precautions
7.9.2,1 from the RVST. Valve TCV-63-1 isola s of RXR 7.9.~2,2 Valve TCV,-63-5 and -3 is l i . SI from the.RWST. o ates both trains of l 7.9.3 Instructions: following: Complete Data Sheet 63-8 in conju 1 nction vith the 7.9.3.1 Turo~RS-63btoopen. 7.9.3.2 Verify the red light on and the g off. reen light 7 9.3.3 Turo HS-63 II) to close. time. 7 Measure the stroke ~~~~ 7.9.3.4 Veriff the green light on and th off. e red 1,ight l (1) it kef**r to the relcrence shecL f or the handsvitch nueber i t, e ~23 J
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sani n ~' FCV-43-1, S. 3 7.9 Oroup 63 8 '/ site puebers: Operatian eWittons for testing 7.9.1 The plant is in cold shutdown
7.9.2 Precautions
l Velve TCV-631 isolates both trains of not l 7.9.2.1..from the RV37. ..sv. y., n.. Valve TCV-63 5 and -3 isolates both tratos of 7.9.2.2 31 from the Rv5T. 7.9.3 fastructicas s Complate Data sheet 43 8 is ceajuncties with the . g'- followinal fars as-63 _(1) to open. . ' 5. j 79 .1 j k M.. S Mi. M*+i= &.- h.. a W.._ ?,
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e (1). Refer to the. reference sheet for the handsvitch number %t^ f.?
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Q. _ e. { r:... ~ ~ F Larr Nobics 2/5/87 O oa es g Acting Plant Honager. SQN l== LA 04bb Q8 1 888 0 p R John Sullivan N 6012 e.a. .s O y S u n'v. PORS. SON subject? " S c_qu o v o h _Eltc 1 c a r Plant (SON) Water Tank Force Interim Spills Recommendat1,ons" __The Task Force investigating the two SC Hanway %~ spills has formulated the attached interim recommendations that should be implemented __orlor to commencing further work on/around the e-~ open SCs. C. Also, attached is the current picture of _t h e -_ Root Cause. We have incorporated comments from both the __ rn e c t i n e in your officcan_d !!. I.. Abercrombie's office yesterday and feedback from the two information sessions that were conducted per _ recommendation # 6. h O .$k
~ _~ v - ~ _ _. x he 2cf 4. ~ ~ ^ \\tt :.. ..rt. *. - - E * "' INTERIM RECOMMENDATIONS OF TASK FORCE TO A1. LOW STEAM GENERATOR WORK TO COMMENCE ) TheseinterimrecommendationsarebasedontheTaskForcereview-to'date I i of the February 1,1987 event with regards to personnel industrial safety hazards. Operations has adequately addressed imediate concerns of the-January 28, 1987, event, by periodic. flushing (weekly) of the sight glass I and comparing its resoing with the tygon hose. The following ~ recommendations have been arreed to by the Task Force as necessary to allow restart of work activities associated with the Steam Generators. 1. SI-166.3 must be revise.d to address the required valve lineups for testing FCV-63-1. and ensuring positive closure of the subject val.ve before realignment of the system. - 2. Retest 1-FCV-63-1 (MOVATS) and repair if necessary (including leak test). 3. The caution orders on valves FCV-63-1, 62-135 and -136 should list specific telephone numbers for notification of SG work stations prior to operation of the valves. We do not recommend placing these valves on a hold order since nuclear. safety considerations warrant their availabity. However, they should have their breakers racked out for the period of time someone is inside (fully or partially) a SG channel head. 4. Ensure all electrical power cords used for Steam Generator work are up to TVA standards and off the floor as much as practical. 7 5. Establish clear lines of communications between control room operators and individuals working on or around the open Steam Generator manways. This communication may be effected by controller's located at a designated control station that in turn are in communication with e-the individuals at the work station. The communication link must be utilized to notify the control room operators of O' start / progress / completion of the Steam Generator work activities and the control room operators must notify the local work stations of all system perturbations, whether caused by operator action, equipreent failure, transients or accidents and prescribe action to be taken (evacuate, clear the platform, etc.). The PA system shall also be utilized by operations to notify plant personnel of such events. 6. Inform those presently involved in the inside containment SG activities of the previous incident. Its root cause, and actions taken to prevent its recurrence. This information session should be presented by control room operators involved in the February 1. 1987 incident and should include handouts and should be documented. As agreed to by H. L. Abercrombie and L. M. Nobles on February 6, 1987, these recommendations are approved and are hereby requirements that will be implemented. ~ 0245R e - /
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' A N. ~ f ~ 'Jg ?- F007 CAUSE ASSESSMENT i l . Based upon the data obtained thus far. In the investigation of the February 1. 1987 Steam Generator Manway Water Spill, the following root cause and action to prevent recurrance has been determined. This - assessment is "not intended to address,all contributing causes to this event, only the root cause. ~ ROOT CAUSE: A procedure (SI-16'6.3. Revision 23) had incomplete instructions in that the stroking of 1-FCV-63-1 was not adequately covered for the situation involved. The procedure did not address the required valve alignment for the condition where RCS pressure was less than 30 psig. The operator, realizing the procedure was inadequate, attempted to'make the procedure work *by supplementing the instruction without going through the change process. The operator stopped the RHR' pumps and isolated their discharge to the RCS. The RHR suctio'n f.lowpath to the RCS was over'lookid, and when 1-fCV-63-1 was stroking open., water flowed by gravity into the RCS Hot Leg # 4 ACTION TO PREVENT RECURRENCE: The procedure has been revised temporarily and will be further revised permanently to speelfy the required valve alignments t_o ensure water does ~' not flow from the RWST to the RCS..SI training classes will describe this event, and it will be stressed to follow instructions or stop and get them corrected instead of making them work when they are inadequate. O e M e-024aR
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y _ N./~MM F)1&d O1l q' !.m [ ,i ~ I.t l f of ( INF MIATION SESSIONS c L, p Unit 1 Reactor coolant Syate= overflow occurrence on 2/1/87; g t ,,g o p) ji. I / 1 .(, t 4 (: . j l7 y a j g, INTRODUCTION OF SPEAKERS ~
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h ',,. ; ~ i I .,.]:.t / a 8 s p.. i,. Purpose _ .lb et ,) d M $' j, g.. ^ . Inform those involved ir. the inside conta1nment activities of the previous ' ( k'k,l8 ' 'y incident. its root cause, end actions 'to prevent its recurrence. "./ hl. 'j') (.0 d Sequence of Events t .p lt<, f iO.! ", ;. i hj
- Work activity prior to RCS spille
'4 .)*j'. i h.' .lg The unit 1 operator was in tlIe process of performing 31-166.8 which was.Jo .f.E 'I l' ' r the increased stroking of several mcde 5. valves. ibis'si-166.8 requires .fl b.Ij,' i that 5'-166.3 cdnditi'nal package to % u, sed for inst. ructions on stroking o o . f k,.i } l' specific valves. The package was issued on 1/31/87 and had,to be completed i by 2/2/87. 3 o ~ '.l: c ,.i, Description of event and activities during the stroking of I-FCV-63-1: c f,.j j,' ' st il .r. 1254E - 1B-B RilR pump was s topped and valves 1-FCV 63-93 and -94, 1-F CV 16, b a: .) -32, and -28 were closed. jil i g f ',*lN..,- 1-FCV-63-1 was opened (opening time approximately 20 secnnels) and then lg I. immediately closed and timed valve (closing time approxl;sately 20 seconds). j 3 '. j 's \\ g; Level indication on T.V. monitor went of f scale and then camera was respanned j to a higher level on sight glass. The unit operator restarted 1 B-B RilR pump i ts !.' It ', and increased letdown to maximum to reduce RCS level. Level indication on the
- 1l sight glass went to approximately,697'2" while the unit operator continued to
, y.j j.;j ; letdown until the level was sta,bilized at elevation 695'9" at 1335E. l f. 1256E - H.P. called from containment and notified the' shift engineer that . 5 l p; 'I.'t : water was spilling from the steam gener.itor (S/C) primary mar. ways. H.P. and c, 1 J, f David Coetcheus called the unit operator at approxImately the same time. r y' r, \\ /; a,1, r ' 1302 - AUO was dispatched to elevation '669 pipe chase to verify 1-FCV-63-1 f ..* "1 f ; '. was fully closed. . 3 i, , f. '. 1310 - The AUD hand tightened the valve approximaccly 10 turns by the manual ,' ), hand wheel. .....~.- 1310 - AUO entered unit 1 containment via scal table room r> v y/// 2,.eec.. ,.,..,'o~ y\\ *" *1 r *r *
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I(,g[ 5. Source of Waters l* M 1 ,P.. T I,' The source of water was determined to be gravity flow through 1-TCV-63-1 1 icv-7 2-1 and 2. 7'o tu //#r s/s d'j. .g N. '! ~ back throush the AllA letdown valves J 4(, [.p'e I ~ 3 Amount of-Water Spillede ~ l {l p 3 f[ fj e.-4.' Based on pocket sump lei,el pump down the spill was.approximentely 3-4000 - T 6 r 5 1 0
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- i ! s e t t one. The indicated leeel change of the RWST was not noticeable.
(y, f,V g -{' Root Caune l [* '8' :., 4.: . e. y l ,)e t.' ., j. ! I ' A procedure (SI-166.3, Revision 23) had incomplete-instructions in that '[
- 38. e' the strokins of 1-rCV-63-1 was not adequately covered for the attuation involved. The procedure did not address the requi,r ed
,v, a lve alignment fqr i,l;yIj the conditi,on where RCS pressure was less than 30 pois. The operator -) p,. j ',~, ' j j reallas.4 the procedure was inadequate,. l.g 9,' p 2, . + / r,' , e. j,
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3;. the instruction without going t,hrough the change process. The operator 48 e. l '[. stopped the RilR pumps'and. isolated their discharge to the RCS. The RllR suction flowpath to the.RCS was overlooked; and when 1-FCV-63-1 was stroking ' l.' 8,,,./.j open, water flowed by gravity ~into the RCS hot leg #4. , j. Corrective Actions to prevent Re c u r r e n,c_e, t l'*.. .... j ;8 '. Approximately 1500, on 2/1/87, Operations Group Hansger, briefed the oncoming shif t control room operators on communtenting with the personnel working on } 'al .e.r steam generator activities and maintaining RCS Icvel below manway openings,.
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4 j e* [d'$,,jl,' Approximately.1800, on 2/1/07, a Hight Order was issued to remaining operators .,i.l *..i r,, cove ring the same subjec t. Attachment "A" .T s ,a m c,.w,a wtos .$,.y. ]. / lo,Q Tf'ik' {.r,Tsue-a,h,.ous have been placed on FCV-63-1, FCV-62-135, TCV-62r136 g s .M rnuto a Ws at.cz hu.*asts it*" to c o n t ac t N s e '*ist. oo y il' f !y.i. L' t bl i i T/ t,a 0,*e44 r,rs. a / .,,, 51 - 16 6. 3 must be revised to address the required valvd '11neups for testing ,> V 4 l I d ' l)j,,,f f ',r,TCV 6 3 - 1. ' ]
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,.,g, t i ' ',I, [,1. Establish clear lines of communicat tors between control room operators and ,.f;.' E j.' k I' q, : individuals working on or around the open steam generator manways. This j k ;;h : )t, communication may be ef fected by cor. trollers located at a des tgriated control ,f k. '! S**}'. station that in turn are in communication with the individuals at the work l? h IgT,' h' f *. ;' s t a t io n. The communication link must be utilized 'to notify the control room ('? -;g ;[,;;,} operators of start / progress / completion of the etcom generator w i (l; ,g.g . and the control room operators must notify the local work stations of all a,..y.p. system perturbations, whether caused by operator action, equipment f a il.o re, g "UE hpptransients,oraccidentsandprescribeactiontobetaken(evacuate,etc.). g 1 , !h,i. The FA system sha11'also be utilized by,0perations to notify plant personnel q q h - ei i % ^ of such events. h,,j ,I n,y,.., s <.g o i, U. w \\ n,,'e...r .e. L '9, .y,.t f;.... i 11 y a.y,, - .i
's/.% p_ ~, y. m a ~.. 4-. / '. 1 G 6 O b s A Night Order will be issued to oprators describing.the. clear lines of k af ' y. ,or around the openl'at.*ae generator manway esowansnicatione tatt teen co ~ that power be remo've4 ' Arom.1-Fcy-63-1, 1-FCV-62-135. 1-FCV-62-136 wh anyone Te~worEins inside the primary side of the, steam senerators. t ~ j , Questions L 'I ) . ;. \\ s.
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~ Pal;e 3 p ; a f \\ ,f *. ). OPEltATIONS GitOllP osi,A30 a V.,, ~* OG/l1/85 ' $g7,. ;. ,~ .'.*).', A*lTAC.HMENT A, {j..;'... ~ ~ s.. .A . *. INFol1MATION Oil TEMI'Oll AltY INSTilVCTIONS 'FOft TilF. OPEf1 ATIONS Gl10U Subjects Unit i Reactor Coolant system f.evel - ~, - i, -* I steam generators. it ~is 1 ~ During the on going modification to the ifnit (.,, I he' responsibility of the OperEtioNis Section to ma'intain the RCS level l !...' t ~ ~. 695' 6" and 697' to prevent over-in'a stable condit'lon between Elev.
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L. _'- flowing thru.manway cover openings on the primary side *, There have been two, events in the last week where water has overflowed. 1.*.* ' out thru monway covers which could have caused injuries and contaminated r- ~ personnel voiking 'in the area'." Honetai.y expense has been liigh due to desiage 'to equipment <and delays jn modification work. V ..l, - l, j M In'the future when making changes to any equipment _that.mtght affeet the f- . j.., level in the RCS. review nnd approval by both the U-l ASE and Shift Engineer is required., Prior to the planned operation notif y'the Combustion. Engineering Trailer at PAX 7102 or 7238 so their people can clear the n.... steam generator platform. L" a. .; a ' s best'not to makE any changes to the RCS, until work is complete in k L '. _ . 'I t i 1: . n..i. l
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~ L-3Qyp-I SI-166.6x he. [] o - N - t /W h. C 'k M DATA PACKAGE COVER. SHEET iL d 4 Page 1 of 2 Revision 21' POST MAINTENANCE TESTING OF CATEGORY "A" AND "B" V .c Unit ~ /' d O / b. M Hode _ [ pg6M '~ ~ f ., Performed B _ ' ' ~ Unit Operatet Date/ Time Started *. 1 9'Nf2 '.2o g o. / Date/ Time Completed a # f > s / m,g-f '
- Package shall not be open longer than seven (7) days during non-outag eriods-List of data sheets attached.
~ Instruction No. .' Data Sheet No. 166.6 ~ Pares Attached ,h. L. 'A 'TC sIf-sa 7.o r? ~ 13 ~ Did all SI data meet acceptance criteria? I t '.no t, Yes No list those valves which failed and Die ~"MR's written. Valve # _ MR # 3,. Valve # MR # N Valve # MR # If criteria were not satisfied, notify the SE who completes the follwing: .Was a Limiting Condition for Operation action required? Yes (exp1_ain in remarks) p Verified By SE _ No (explain in remarks) Date Reason for test: Time Required by schedule ___ / ~ Maintenance complete on k)Ad2 7.lO& 7/ g Another system (._ (Instruction _ 3' _ Plant condition (explain) ) inoperabic ~) Other (ex ...................... plain) _ q-Review of Tes.t Results ~ b Date 2-y O Time pro I ASE ~~y Review and Approval of Test Results 1. Does any valve require a change in frequency?Yes No _ 2. Does any valve require a return to normal frequency? Yes No N/A f/alve(s) Affected 8-4
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- Does any valve pquire a. change in maximum allowable stroke time?
Yer No-~ 7 Valves affected and,new maximum allowable stroke time. ~ ~ ~ e e 4.. HTI-14 completed Ahf4fdt.f] ~ 9, (, Engineering & Test Mechanical Engineer ~ ANIT Review of Test Results l-ANII Inspector - ~ Date Remarks: l s~ e QA Review of, Test Results ~. e QA Staff Date hemarks: 0-e e f '8% e 4 4 l l l .1 s G 9 C
b-e. ,:.~ ~. .-- -...+- ;--.- ( 6 l SQNP SI-166.6 Data Sheet i Pase I of I l.evision 21 POST ttAINTENANCE. TESTING OF C$TECOHY"A" ANI) "h" VALVES ~ ^ Local Timing. TE.ST Per: (1) (2) ( tj Reipsircil RO/SRO Inst. Inst. Test Complete V.alve /ttR/WP Date In; t i.il' ' YES/NO Ini t i.il Si s. ';t ep Page Initial /Date Tc 7-l"If'2L-) 0221% '2 2.:.hl7_..bGW-W. f'.@ ~ MYN 2 +/).. m .e ..qa e. .e g M_ - y s. p.a ...g... um.
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g- [ f } l. ' ['[, ' Standard Practice Attachment 1 ~ ' Page 7 1.J }, Unreviewed Safety Question Determination SQAttg J' -fC, M. j( 7 - / l 7 Revision 6 },h. Change Description J l-11, /s, % k.co. 24 0 . b e.fus, 1, a A A A s ks. r L ao n A A & J. 1% d*> FLU -t42. I Lwn . "T! W w.,'M 4A > Rs ,Do's.&. h s 0 M'bs. GD A n,.,., 2, q M Q*' Y. ' 2/, nht),. s u ' fx.,'._A A L .L*R JL Abl2h ~% w&#. 4240 OvUuG W && HM j. Check block (s)-below to describe what is being evaluated. fcv.(,3.(,, System, component, or structure being changed is described in the TSAR F5iie reviewed procedure (Including temporary changed ~' x-- Special test or experiment Special activity, abnormal conf.iguration, etc., which.is deemed to need a USQD 4 C. Is the probability of an occurrence or the consequences of an accident or malfur.ction of equipment important to safety previously,qvaluated in the safety analysis report increased? YES NO ' X c Explanation: AGA. Me,bf m D. Is the possibility for an accident or malfunction of a different type tha7 any evaluated previously in the safety analysis report created? YES NO 4/ Expl.natlon: eAp> e M m /_, 0 a. e ~ e E. Is the margin of safety as define,d)n the basis of any technical specification reduced? YES NO A__ D Explanation: ..!Lu n D'-_,M ['
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- Note:
PORC and Plant Manager signature may be waived if change is to non CSSC equipment, however SE must sign in space provided - otherwise N/A. 1 .v l e e
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SEQUOYAH NUCLEAR PLANT [ Sur.VEILLANCE INSTRUCTION SI-166.3 FUI.L STROX1NG OF CATEGORY "A" AND "B" VA!.VES DURING. COLD SilUTDOWN Units 1 and 2 ~ ~ ~ ~ Revision 23' PREPARED BY: McPherson/Terpstra RESPONSIBLE SECTION: Engineering s%. REVISED BY: - m Don Goodiw/ S'JBMITTED HY: a E_ Rpsp6nsibI'e Section Supervisor I' ORC REVIEW DATEI APPROVED'BY: C' Plant Manager ~ ~ MAY I 31986 -DATE APPROVED: Kl. c- .i D j Reason for revision (include all Instruction Change Form Nos.): .+ Revised in accordantc with ICF 06-0675, PORC reviewed and l g, approved 04/15/86. l l-l l The last page of this instruction is numbu.r: . 7. 4.. j { ~ lu i c .~ l l .I
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~ -~ -- SQNP ~ $1-166.3 - tJnits 1 & 2 j 6 0 ~ Page 1 of 1 Revision 19 ? 7.9 Group 63-8 Valve Numbers: TCV-63-1 5, 3 t 7.9.1 Operating conditions for testing: The plant is in ' cold shutilown 7.9.2 P.ecautions: 7 ' ' ~ ~ 7.9.2.1 Valve l'CV-61-1 it.olaten both trains of RllR '~ , from the' Rt/ST. 7.9.2'.2 Valve FCV-63-5 and -3 isolates both trains of SI f rom the IN."iT. ~ ~
7.9.3 Instructions
Complete Data Sheet 63-8 in conjun tion 5/ith the ' following: 7.9.3.1 Turn 11S-63 _(t) t o open. kerifytheredlightonandthegreenlight g 7.9.3.2 off. 7.9.3.3 Turn llS-63 _(1) to c!c r.c. fleasure the st role time. @p ~ 7.9.3.4 Verify the green light on and the red light
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h s m.mns.=-m.J G.~L.1==.- - A .. s. m...... L 14 aP 1e B SQNP SI-166,3 Page 1 Revision 0 Data Sheet flo. FCV-63-1 Page 1.of 3 ~~ UNIT _. ./ DATE .7. Y,-J 2._._ PMT-:, V,,,,, yr.S No. / MR nn. f.)) J O (,2 Valve Number: TCV-63-1 Valve flomenclature: fiWST to RllR, pump Dde N Saf ety Train: A Maximum Allowable Stroke Time,: U1 = 3od sec.. u2 = fo.q..sec. Maximum Allowable Stroke Time Determined By:. Pre-Op Timing Position: ettr: red C l.oS ED h!f dl fN Drawing"Ref e'rences: 47b'811-1, 45N779-33 1.0 - OPERAT!?JC cot lDI,T_ IONS FOR TFETI!iG The unit should be in cold shutdown. s. 2.O PRECAUTIOt!S ^ 2.1 Thas valve isolates both traips of R)(R from the RWST. zu 2.2 Refer to T.S. 3.9. 8.1 (mode 6). Observe RWST level during this test. B RHR may be stopped, if required. Observe !.CO 3.4.1.4 (mode 5). tv 3.0 INSTRUCTIONS
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3.1 Turn HS-63-1 A to OPDJ. Verify the RED light Oll, GREDJ light OFF. 2 Versfied Dy: _ h.. i. O!62Thlt s
.- x ,q l' ,hacl50Yl A ~ SQllP SI-166..$ Page 2 Revision 0 l .l Data Sheet No. ItV-63-1 pago 2 rif 3 3.2 Turn llS-63-1A to CLOSE position and measure the st'oke time. Verify tho GREEN light 0% ;1ED OtT. Ve r i f i cel lly :,, ,[gq / 4.y'.py
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/)./'.f/ U0 or SiM RD40TE LOCAL - i t'. I1 _Sec. _,,_,,0e l') Sec. Stopwatch ID: ( f3 Sty;.u.it ch ID:. ,d [ 7 f y'~ CAL. Due Dato: @ y.7 7 CAI... No Da t e : f. //. ['7 N Recorded BY: [// [,fg (f(,dhgj,,tjf 3 C, sa-MAST: UI -M Sec.. U2 32h2 3ec. 30 0 -).C 1,..... o -. QMDS tracking log updated. LIO Date o. tJO[E: If valvo cannot be Icft in the Cl.CGliD position state reason o. j and "As Left" position in the remarks section. Record in the e configuration log. 4.0 RO'.ARXS. 0162!/.,1t l
~ 1. ^ Q J)(, ff I] SQtlP o .SI-166,3 Page 3, t Data Sheet No. FCV-63 ! ~ Page 3..nf 3 [ t ' 9 )
- 5. 0. AC.CC. P.T.AN. CE..CRI TERI.A.
.t'! P . ;); i ,( s A. Did eit hnr the remote or local ~ st roibt t am's exceed ths MAST.
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- ^
Part C of SI-166. +. Evaluateel 3y:. _ - / = f,10TC1 This section is to be completed tey e..a_n F";i : -: ; "1 Test th Mc chhio.\\ - S Chh. C. Did the remoto stroke tima exceed the ^ CMAST of - ~2G.le
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~ ff . ~ - W ; ~~. p g h M f} h.;' l.(; n:a go-'s?:17:22 STD. FCACST HILLS ' 7' ,y P. 2 - ],j[* N If . /[ ' A ;F Westinghcave Bafemmos No(s). ; 5 '{ NS-str*c}2g nevinion'i HS-RCSCfrc/IrB7 bio 4 nev. 1 NErrDREM2 }i NOCIQR SAHTf INAtlATICH OE3C I.TSr n 1). x ,(- 3,tc::An f::.:::q m cx _a_ -,1 + 1 + d ' l A) 00:0< LTSr APPLICABIZ 'IO:Evansatlan of Fgtgt, 312 Fi*n Insulation A ji n.j c.ii
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ny Mimirer in the_ heter Cboinpt Systm a. a ~/ p ' ' *s xy a' a ,.L j ~ ,,l[j, . 3) *!ho safety evaluation of the revised gwwhwe, design charge or i ~ ' Mfication recNIInd by 10Ci"do.59 has been prtpared to the extent recpired'eni is;attachad. 'u d mtety evaluation is not required ar. w, in incx:cpleta ter any m1,,eg:m.in an P2pu 2.1 Parts A and B of this Sai'ety Evaluation Od List are to be ccx:pleted W anly on the bacio of tha safety evaluation perfnemaa. OE2C LTSr - PARr A 1
- 'At f.
- (3.1) : Yaet_ W1' A change to the plant as rWr ribed _in the 13AR? A. (3.2) YrA,' _, No1 A change to prnwhwes as dW in tbo ISAR? (3.3]L Yess 'No l 'A test or,axperi;cnt not A m ibed in the FSAR? (3.4)..Y W - ' Nql A charrje to tho ' plant technical specifications = (A,ppardlw A to the Oparatin) License)? v,. , n.. ?, J. 4) QIECK LIST - PARr B '(Justideation for ParkdD artnera must bo included y on ' pap L) ['j No1 Will thh pr,ct.nbility of' an accident prt:vicusly a t c .(4.1) Yea
- ~
evalud.ed Jn the ISAK W.4,, Yes No1 W11.1 thd rhnsorguemes.be ircreasedt of an accidant previcualy i evaluou:d in the FSAR ba <incroat.ed?s
- f. '
(4.3) Ym _ No1 May tho possibuf,ty of an accidant 1Alch is l 1 difforent than asy alznady evnlustad in the 13AR 5 i be cmted? ' i (4 4) Yes, No X Will,thu :xtability of a talfunctica of eqii mmit s ~' t important% sarcty pruvini'ly evaluated in the
- j.,
ISAR }4-i n n.'ard? f -[ ( t,4,5) Yes No1 Will t 3 conccquenoco of a unitunction of ec ip-2 <iL tent ingqtant to safety pcuvinmly evaluated in tho IBAR be ie*? x y (4.6) Yoo Hol Mayythe pxcibility of a ralfurction of ocpip: ant a t' irportant to safety different than any already cvaluated in tha ILAR be creatod? F (4.7) Yes No_X_ Will the margin of safety as defined in the bw y to any technical riccification bo zwhma? Y I l 1 r' , Page 1 of 4 I r op j c. ^ /.i f 4 x i y ys (t f/ I L k'.
p ov. era 20 *07 17:23 STD PCPCST HILLS P.3 Mc 2 cd U 'If tho answarm to any of tho above questiens a212 tinknown, indicato tznder 5) RD9 JUG ard explain below. 1 'If the anawar to any of the abcno quest. ions in 4) cannot be amwertd in the recptivo, basai on written safoty ovaltation, the charge canrot be appzxnod without an application for license Eu.ahTt subnitted to tho )GC patuant to 1(V.2R50.59.
- 5) RD'Nus:
NCIE 'Iha followin cvaluation, g'aimrizes tha justification t$on the written cafety ~ (*) for answero given in Part B of the Safety 1v. aluaticn Ctcck List:
- * *
- ATIAC ED * * * *
(*) Rnferunce to dm~nt(c) containirs written safety eyaltuticn:
- ArrAC ED * * *
.M M FM N Section: __ Pages: Tabico: Figurm:_ Paascx1 for / Doccription of ChangA:s sb%s LL - by m cer % 3, v4
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- teswith aerme d s/ W M Kt # mee, shon V s ceordimtas czum Fanscr(s):- W . w. mramt rate:_3/_v /D. IArlear safety croup Panagcr: && VM /f
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- WA 20 '97 17:24 570 - FCPEST HILLS P.d o
,$c ]g[/\\ p. MS-SEC!r67-120 Davisicm 1 NS-HcucirqrIre7-404 Revisia) 1 brmerncerCH _ ~ D2 ring a recent cutage at Sequoyah Unit 1 some parts from a devKd Steam Concrator U-Berd Hoatar wara lost. Die loose parts iicntiflod in a worst casm situation wru avalur. cod in Westinghcuoc Safoty Evaluation SECL ccrsmic pallets evaluated in tho aforementioned evaluaticri, it ard n2st bo pm-d lost in the 'neactor coolant system (RCS). O evaluaticn vill azaass the ability of Sequoyah Unit 1 to start up ard aantinuo safe plant operation precu:nirs that up to fort 1Wt (48) inches of 3 dia:n/16 inc:h diacoter Nextal incult. tion and 1.50 inchos of 2/2 inch 12.44 gra::n of Nextcl.' 'Ihis cx:nservatively large acount of Ncxt ~ rgrtments an abooluta worst caso situaticn. EVAttWrigj into its fibrous const.ituento at pri:cary system tcIt has been detcm As such, the Nextcl vill not be a cc:rurn fran a loose parta point ofaporaturn and prxmrc view. 'Iho Uextal'has bocn intzrrhvwi into the RCS in tm ways, the first ~ is frem the normal process of utiliziry the heater prebe. As tho prt trnnxt within the steam gemrator tubo scs:e of the Nextol is lc t d e is the Nextel was sicply lost when tho at9am rienerator bcul w a ue to N vator and sooo of the insulation was in the bowl. 211s'evaluaticri vill dchn into its chemical cc:::ponents. 'Ilva follcwinry tchio repzwents Bav. 4, Tablo 1.5mavi at:ount of each erw=rd allowed in the RCS, as given in SIP 51 s PAnn"Erct cemte <n<mtmo MW. ctNc. 'IUrAL Esk 1m 9 Sus.Sn1 W 5200 Sun. Solids 0.2 40.0 S11 % $1000 sio 1.0 240.0 2 Alu::dt:u:n $50 Al O 0.0345 .22.67 23 Magnesit:n $25 ftf) 0.0414 9.95 Page 3 c4 4 e L
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- '. tw -20 '97.17:23 -STD JY#1'37 HILLS P.5 h()
F NS-fECIe67-120 Davinien 1-NS-RCECIrc/ Irs 7.404 awvisicn 1 o ~ ~ c.a fol.lcuirry amounts of each cxxpound have been inWmi to ths pri::nr/ syctan baand on a wonrt cace cituation. 'Ihose amounto covne both the ceramic pellets ard the Ncycc.'. insulaticn. .cem-Ie Pn11 eta. Abruided Nextn1 rerrt t,*metM sus. solids 4.so g 4.11 g s.33 g Silica ~ ~ ~ 3.11 g 1215 g 2.32 g Alumina 0.22 g 2.88 g 5.81 g.. Magnesia 1.33 g o co g o.co g As can be seen by the above table, the total amounts of each %-d.: g Suc. Solids ' 17.24 g., Silica 6.58 g., Alumina 8.91 g., Magnesia 1.33 g., ac+mily rtprocent sem11 a::cunta of the alle.rables. 'Ibe above t&ds xvpresent apprax:irntely 97% of the suoparded solids tren the pallota and. the }kuctal. 'Ibe ru:aining 3% reprosenta various traco substances i.e., g 1%0, C::0 and I'ec. 'Iha totala for suspended Solida do include the listed arounts of S10 ' ^10 ', and Mjo. 2 23 Ihsed on this evaluation, the et= (ra1 effects of the ccrsmic pelleta or the !koctal insulation does not Impresent an unresolved safety quccticn trder the requirements of 10CFR50.59. O* p e a Ibgo 4 of 4 e
.g 7- ~ ~~ ~ -Fi3 24 'i' 20:20 A-si:"';T1;f! Ajit.0; 5d ll e Feb. 24,1987 h I amel subject: Response to NR TVA 487) 00023, Unaccounted, .Heator Beads t The attached Safety Evaluation Report is provided in response to NR A (s?) 00023 concerning unaccounted heater beads. The assumption that.all si of the nissing beads are in the.orimary loop'is a worst case condition. Tho raport responds to the five opocific questions pesod by D. Coetches,(TVA).; To elaborate'on the fuel question,-although.it is considered an-incredible event unsupported by Westinghouse experience in the evaluation c f foreign; objects in the RCS, if all six pellets congregated in one location l - adjacent to a fuel rod and formod a soft material, such does not pose an u unreviewed safety question and does not precludo safe. operation of the plaqt' The detailed Chemistry-Technology response is included in this j - transmittal. Thora are no unresolved safety considerations related to NR TVA(87), - 00023, and the situation is acceptable as is. { D.R. Stener Principal Eng. PHS Eng. R.D. Burack ,( ^ Mgr. PHS Eng. cc R.M. Clark ~~ C.W. Hirst C.G. Elder 3 J.S.Galembush B 4 4 6 Y e. 9 m 1 L
h: rts.Zb87 13519 R".D 701 R.DG 412-2!6-ti7 43. - P.C2 GC C hO I' a, n.v4.4-1 .( ] \\. W 7-394 nav.- 1 Westin@x2rsa Mereros N(s). l= gg 30 CLEAR SAFEIY EVAu.m u i CECK I2Sr
- 1) 70C12AR FINir(S). BZD0th.V DL" 1 l
2). CsECK 12ST APPLICABIE '!Omvalt*st4m of Sir f 61 caramir ps11ots - ]j stinnim in t he r-r wie th1M h 1 I 3)- Ubs safety awaluation of tbs revised p.- twin, design charge or acmii.ficrtic:n reg 11 red by 10CFR50.59 bas been r % M to the exterst regaized armi is attu+=4 If a safety evaluation in not required er i .is inec:pleta for any roamer 3, glaM cn Phge 2. i Parts A and B of this safety Evaluaticn C nc$c" List are to be m plated -{f cnly en tbs him of the safoty evaluaticn p=IQ. 1 1 i cszcx 12sr - mar A 'I s.( (3.1) Yes No1 A dange to the plant as described in the FSAK? (3.2) Yes, lb_X_ A chargs to r' 4wto as described in the FEAR? ( (3.3) Yes No1 A test or expsriment nct W in the PSAR? (3.4) Yes Nctl. A change to the plant tactinfrm1 W eih (Appandix A to the Operatir.g Licaine)? m
- 4) GIX:t 123r - PARr B (3Lstifiestien for Part B answers Eust be irrltriad cm pags 2.)
,.j (4.1) Yes,_. Fo_'L Hl.d tbs prctability of an am4M previramly evaluated in tto FEAR be it:2maar17 i e (4.2) YedL_._ No1 Will the e-- rwicms of an M M penvicusly evalustod in tus FSAR be inczmaa't? (4.3) Yes_ No X Hay the pcesibility of an amid-se Milch is di.ffatune tbsn any already evalusted in tts FSAR be created? (4.4) Y::s_ Ib1 ; fill tts.,#_' -k4'dty of a rnifuncticn of =dM in,wLud. to cafety prWreaty evaluated in tta FSAR be 1,u J I (4.5) Yes_ lel Will the wwmicos of a inalfuncticn of ocpip-t Derit i::pertant to cafety previrimly avalust.'a:1 in ths FBAR bs ircreamed? (4.6) Yes_,,,,,1101 May ths PihOlty of a tali \\:rcticn of @im iwLant to safety differtric than any.airceay cvalusted in tE ISAR ba c::reted? ( (4.7) Yen _,,,,_ }b1 Will the margin of safoty as daffnsd in the baans l l to any technir=1 m=< i f b aticn be rodoed? 1 l l Itga 1 af 6 1 1 i
p.c3 ~, ' TEB.25 '8 '13:20 R*>D 701 R.Dr,.312456-6 t3 u - t. ,.f - - - ~, - g 6 7d f tzdar 5) REMMp3 and mimin W. 5 fib If the amuur to anlf of ths 'ah::m cy::sti:r::s in 4) comet ha artvuxud in e. n uo, b d on wie, f.=y -1u u=>. 22 % be aqueved vitixut an m14e tien far 11cese amuruhant submitted to ..l 12m MRC pursuant to 1DCTR50.59. ~ l
- 5) =7K3:
M -n.- 2bo folleving r==rizas the junti.ficaticn 1:pers the writt.21 mafety evalustien, (*) for answers givun in Part B of the Safoty Evaltaticn Check List ~
- p r=wn e***
(*) Dafertrian to haner:t(s) containing writt:::n safety evaltation:
- ATE *LSIID *** MD CIL'MISTtt - m 6.3.1 *f cS8 'JITU CS8/ c 5r3 e
72 PSAR12PC E e ' - 1: Pages:
- n hi ca:.
Figurce: he far / Dascriptien of change o 10 FSAR UPCUE 3PENII2D 4e w J/#f IN Pre l;nrad by Obclear Safety): J.8. GLY m t Date: esL Ctaniinated with Zlnginaar(o): J. P. IMAN Dsta:.ZjZJ 67 Cbcut11 rated cn:up Ftinagar(s): 6. ? Date: 87 cate: 7 9 C2::c:.tir.sted with Engineer (s): N I.Er$ E--, 2.t.'/f M.ated c.t:up :anager(s): c1 M% c,t..>Ac/r7 L c= = cm., ~ (m:
- c. x g/w J./lf/r7 W vr 2f:ty crt:tp :Sn q::r:
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- 1. mmr
- cate:
rarp 2 et 6 b 4 e ____-___m________
-FEB.21 887 13:22 P b ~01 E4.ts; 4th2'4-67d3 - - - h.b5 G O f' h MTrr,.47-088 FM. 1 l MS-RCSC2rc/L-87-396 Fev. 1 1 j i .s SICLtnaW t. WIT l'ERTIY EVAUATICI FC*E*.Of Ca7IrIIS Di 'DE RDL~It'R CD3IRtr SYFI1:M l
- D,xrirq a recant cutage -at Sequoyah Unit 1 a steam Canarator U-Bard Heater Anaenbly Fw wack;cd insics a otsa:n garwrator tubo. Upcn ziceval the beatar==ely came apart. All parts of the heater were accxunted for excnot for six (6) cenanic pellets, mami cm the nest ceriserative, worst casa situaticn, it is==M that all six (6) pellets were vashed into th Deactor 0:clant system ard aust rxu be censidsrtxi lec:a fertign
-j cbjects. 'Dtin Safety Evaluaticn vill assess the abi.lity of Sequoyah Unit l' to start-up ard contirus safe plant eparutien =-ming that the six .pelists, reu faraign etrjoces, azu misairs are unrecovazid in the as. 4 2.2 forsign cbjects hsys ths folicwirg diz=mmicns: l Dismatar: .330 inctims f .-(. Imrgth: .290 in:: hen Maight: .80 grz=s oni::h Notes 2.ars is a 0.114 inch dia= star holo in the contar of cach pallet. gs ( C.s fcznign cbjects havs ths felicuirg chsmic:11 v iticn: Hgo ~ 27.8 waight parcent i 64.8 weight parcent B10 A1 023-4.6 weight W 6 1 Ard tr..ca c= cunts of the follevirg ouhstarcess F.a2, C::0, 2::0 0 IC:UT :PARIU ""'.W.CJ:Cl:
- Fue.1:
,( 2 caura of ths size of the foreign cbjoc s ard thair 1ccati:r. in ths ycs durirg cparatien, it is highly un.;,ikaly that thsy cculd :aka their vay Page 3 of 6
voua uv emu um m8 EM aM-2%--Bod ,:g cge )di c grr*;-817-oHI Pev. i 13-Mct#/L-d7 595 9,=v, r_ ( into the tual aa=-rh11es. However, if ths feraign :bjects usre able to r.ig: sts thr:.2;n tha rsac:r ~1 nt ayaten and into tha Icwar plenum of the reacter, thsy oculd pessibly e mtrair.ad in tha emlant new ard sub +quantly becces entramod in the fuel to: tan rozzle arma. Infor=aticn and di-~wcm partinant to saca ccrriiticn azn given talev. Natarial Within the Fuel A=== bly 'Ibn forsign objects ocnole.ared have a size such that they oculd not pass thrcxx;h the botten rozzio plata unloss significant voar or deforsatien to the parts occuru. If piccos did pass threugh the betten rozzle ard thrt:: ugh the lowar grid thsy wculd not affect the DG evaluations for this Tarts en cpan lattics fusi a==611as fixiicata that a bicckage of coz,z. up to 41% is acewtable, with di erarco of the stagnant zone behind ths flev biccksga aftsr 1.651/Ds (length /ocpivalent hpirsulic dbr-atar). '!hese types of local blockages have little affect en subchannal enthalpy ' rise and offcet minor parturtstiens in local tass valecity. - In reality, a Icesi new bicekage is sog: acted to pw.a turtulonen ard thus, uculd lihaly ret affect om at all, c=sidarirg cne d4Te itien of the pallots t.%1ves and althougn it in cens44 md an Au uuiblo event, ur6tod by Westin:;ncuso axesriones in tho ovalustien of fornign chiects in the a RCS, if all six (6) pallots -jm;rotod in ens lecstien adjacent to a tual "{ red and fer=:nd a coft :starial, such dess rot pese an unr:rviewed safety qucaticn and vi.11 n=t i.clude cafo cparatica of tha plant. Material T. r::pped '.:y S.a D:::= 1'.c::13 Pinto B Dec:suse of tha oi:s of ths flew hoics in ths betten rozzle plate, none of tha fersign cbjects vill pass threc,. ths bett :n rczzle plates and up 5 Anto tha feal assa:daly unicas significant wear er d.afarcaticn to tha parts cc==c. Aamnig, c=nservativaly, that tha feraign cbjects can pass Qataly th. :gh ths reacter c=ol:nt aysten ord be t.neped by tha betten rc::les, ths flew to any em as.oe=bly v1.11 ret be cx:cplately biccked off. Miltiens11y, ccc-IV predictions Jrziicato that ovan when bicc6:ya cc=:pletaly covars ths r.c::lo, full recovery of flew occun abcut 30 inches ~ dc'w.a:s of ths hice.'c:qs. 'Ihus, inlet bice.'egs effects wculd b3 lir.itod to tha 1cuar pertien of the cctivo cero, uhsro CG ard IrC\\ ara not lir.itiry. Roaults of 9.s thar=al-h'#a.11c cvaluatica chew that tha presare:2 of th::ce fe:mign cbj x:cs is not _ua to result in any cparaticasi prcbims frc:s a flew bice cgs /I'm ctand;cim. Cnt c=rcorn !.s tas pcomibility of edditiensi dabris bainq generataa by cava:=nt of tas fer ign cej: cts thin.x;ncut the reacter ccolant systen. If the additienil debris aro u=all arce;n to p.iss thznx;h tho. bet *4:n rcz:le ficw hoics, there is a potmte.a1 for ft:1 L ccir:q failures to cecur. Ms has been cc:cnctntad at other p1N ts tusro dabris hsvu % ledged at ( grid ctrspo erd caused ens or tora frutting failur:s. Houover, fcal b failurus et this nort do rot cer::titut.s an u:.rtvicwed Ostety gesti:n. Page 4 of 6
.--_ ~ _ C O { mr h q7-esis t w, i NS-PtScircMr.h t-D5 Pev 1_ opad clustar Centrol Assenhly (TK5A) cperahen: ~ 2.s p:t.::ntial fcr the forsign cbjocts to advarse1Y nffect RO:A cp:xrsticn is extz: csly starots. Tcr BC:A cpsntien to ba affoc sd the to:lign cbjects wen.'M Mve to entar a guids tubs th:tugh cne of tbs fic:a openirgs in the Iram guids tubs aria. Cncs inside ths guida tubs the feraign cbjects wculd have to tme betdeen the red channals into ths centar of ths guida tube and rise up to the top of ths continuous guida tubs rugien, onom abcne the contintma regicn tha ' tbjects wculd than have to crient thscoelves in auch a way as to lay acr=ss a guida plata and block ths aph channal. It is canaidared an auctr=saly receta pcosibilit revecients and pi% r y that the feraign cbjects could affect all the - q to affect preter cperatien of the RCI3. ,If this vers to occur, such a circu= stance does not ecnstitute an unr=nolvsd safety questien sitco analyses have nhown that the reacter can ,ba safal vit. % y shutdcen with tha highsst worth centrol red stuck in the fully pcsiticri, m
- Rsacter Ctolant R::ps
[ It is expected that ths rtactor coolant p.=po vill bs ur:affo: tad by tha I i foreign c=rjoc:s. in y,c:p vibraticn d.aractaristica er inc:: case in lec.tod reter accidantW.s precebility. ccesidarod to bs unliknly.Significant 2:scr.rtical d==ga to the pu::p i:peller is )
- Staam Canaratcr If ths fcenica cbjects vers to entar ths otcas g:narater chann21 head, the offccus wculd be poenirg-type impac en ths t W-r1
- W ard channel head.
Pedal 51 steam genersters have p:trtiviira tubs erds abcVe the { N N cleddirq. minimiss the pot =ntial for a prir.sw.macniary inak to ccc:r.'D'.s integrity 1::;:e.cdrg offects of ths fersign cbjects en thm channal h::ad, t-W 2.e cleddira, ard tuba onda aru exp:sted to bs negligible dcs to their mini al mess. ) C.2 ternig. Obj= In M U-borca dran to ard includirq rce 1..r.=h to that they can pana thtur:h all vp to 1.9 ire.co in leryth can pans thrt: Analyses have shcun that en cbject incit::iirq Ic's 2. Lgn all U-butis dcwn to ard I:: pac-?rg cf tha ic:aign Obj :c:s cn ths 1.3idas of ::t$a::: q:ra::r. ::,,.c 3 is jt n d not to significantly affoct ths integrity of ths tuoun. 'hme e vall thi.trifrq is caly an insca staru locoe parts cculd h ledgod are cr:ats %cr: sar et a lecali::ni recien. I 3 l Page 5 of 6 ) u
78 r.d es 0 s s.t.s c.,y.vo so qcy.%. ? q e lidl! l mY"-M asi!I %v. L [ m =ccco cm-u-396 nev.1.
- fftRS 4 cas
. v ru : :.c. Valves 'Ihm fcmic;n ctrject.3 wi'd rpt pr=verst,tb MRS valvse frem cg:enirgt b.ft. Sirca tha ERS - they i:caald preclude TJWi valves fren oiWi.::; :=rm1'2taly. Varificaticn of cles.tre of the RHRS gat.s valves is aernp lankage tasting dLxrfrq heatt:p follcwirq rafualirgs. ' Heat Deharger tw foreign cbjects a:511d entar a resh'imi heat rancwal heat me:harger. If this were to c'm a tube loak, it..would be identified via high g radiaticn ard high curva tank lovel in ths %w.c coolirq water systa:n d cut for repairs. ,ard the affected heat exchargar cculd bs valve Pu::p. s tQ p th9 fC3*3 @ Cbj0 Cts With T'D p W.s possibility axi..ts tMt oc=e part of a p.=p 4--allor cculdIt is twt believ are a D.S M pna.-g, seizurs. km chipped by an iz: pact trc:2 ths Icesa part. would impair p.=p cparaticn. .. f(
- Paacter Irf.arnals ard Vassal:
a. ars ' twt e:q: cted to af foct alther upper or Ecuar W.s fcraign cbjacts,D.s mes of the cbjec:s in rat c:nsidared cufficient Fretti.g ard reacter intarnals. 2, to 12:part any signific:nt i=;:scu 1 cad en roacter ints:nsis. wra is air.o voar. dr=n tha feraign cbjects en ths interr.:Lis' w c.u.. % to be imignific::nt.- C'&%f witien of thz famign Ds. sed en ths pr:nicualy cantic=ad cWm1 e'r ebjects, the offocus en pri=ary c e atts chculd thm cbjoeta M~1ve ard ~' go into polutien are e od to ha draignific=nt. C:::lCII*3:C;: 2.s ability of Squoysh thit 1 to start-up srd centd :ca cafo plant 1cn ebfocts in ths Imcter ccolant cpra' 1:7. ri.h th3 rif. .ci-.cd fer:x;.: c ;m:.*. n vith the fer:ign cbj :c.s in uyst.:.:n C..::.:.a ,:.u.; f.:.the roacter c:clant :::yst=n dcos ret raprenant an un: 2 Viewed o&f-ty gaostien.'. k. Page 6 of 6 4
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i o ATTACHMENT 7 TO ENCLOSURE 1 1
~ 81050600777 vv4. y .n -..s., eu,rco 3,rm. comun.sr 001 '87 o50t 800 Mem 'andum TENNESSEE VALLEY AUTH0juTY V 70 - 3. A. White, Manager of Nuclear Power, LP 6N 38A-C FROM s' R. K. Solberling Director of Nuclear Manager's Review Group, 716C RB-C Daft
- May 1, 1987
SUBJECT:
NUCLEAR MANAGER'S REVIEW GROUP (NMRG) REPORT NO. I-87-01-SQN; SEQUOYAH NUCLEAR PLANT STEAM GENERATOR REACTOR COOLANT SPILLS The report of NMR0's Investigation of SQN reactor coolant spills of January 28, 1987 and February 1, 1987, is attached. This investigation was performed following-your request to me in early February. This report discusses the factors that contributed to each of those spl11s and includes several findings to help guide management response actions. In several cases NMRG found that the substance y of the findings was already known.to plant management and corrective action was in progress. For.several findings, however, including N those that were not direct causes of the spills, additional corrective action appears warranted. g NMRG personnel will be pleased to discuss the findings further and to provide assistance in correcting the problems noted upon request from you or the cognizant managers. m l 4 ? R. K. Selber11ng T RDS:RDC: PAP ! C' Attachment cc (Attachment): O RIMS. MR_41,12&.-C p* H. L. Abercromble. ONP, Sequoyah C. H. Fox, Jr., LP 6N 38A-C R. L. Gridley, LP SN 1578-C W. H. Hannum, BR IN 76B-C C. C. Mason, LP 6N 38A-C L. M. Nobles, ONP, Sequoyah 0555U f., L Buv l'S. Savines Bonds Reeularly on the Payroll Savines Plan - _ _ _ _ - _ _ _ - _ - _ - _ _.}}