ML20062E079
| ML20062E079 | |
| Person / Time | |
|---|---|
| Site: | Salem, Indian Point, Arkansas Nuclear, Surry, North Anna, Duane Arnold, 05000000 |
| Issue date: | 07/29/1982 |
| From: | Michelson C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Lafleur J NRC OFFICE OF INTERNATIONAL PROGRAMS (OIP) |
| References | |
| NUDOCS 8208060327 | |
| Download: ML20062E079 (17) | |
Text
.
Sc! - 2 9 ?
'Et l
2SL E 2.5 JUL 2 91982 3 11.1
.M7 MEMORANDUti FOR: Joseph D. Lafleur Jr., Deputy Director Office of International Programs FR0!i:
Carlyle Michelson, Director Office for Analysis and Evaluation of Operational Data
SUBJECT:
IRS REPORTS Please forward the following enclosed IRS reports to Mr. Otsuka of the flea.*
1.
Spill of Contaminated Water and Contamination of Personnel in the Auxiliary Building.
2.
Failure of Control Rod Guide Tube Support Pins.
l 3.
Loss of Reserve Station Service Transformer 4
Large Unnoticed Leahge of Service Water from the Containment Fan Cooler Cof1s and Supply Piping.
5.
Cracked Ilydraulic Spaed Control Cylinders on Pain Steam Isolation Valves.
6 Spurious Trip of a Generator Lockout Relay Associated With A Diesel Generator Unit.
Original Signed by Carlyle Michelson Carlyle Michelson, Director Office for Analysis and Evaluation of Operational Data
Enclosure:
Distribution: f-As stated Central File /
AE0D Reading File AE00 Chron File D. Zukor, ROAB W. Lanning, ROA8 K. Seyfrit, C/ROAB C. Heltemes, nD/AE00 C. Michelson, D/AE0n DN 8208060327 820729 PDR ADOCK 05000247 S
PDR n
sw
?
- oma, ROAg...
C/R0AQ A g..
- wn n M..............
ROA'3 suR-Mo..nz.u.k.er..t.h9....
... W.L.a.n n. ins.......K.S.e.y..fr.i t... !.Wi!
mes.
.0111.chelsm....
....?.@..!a2......... z/.g48.g.......z/.2.,La.g. 1.4. ag..
.2/.h./.aa.....
om>
NRC FORM 318 (10-80) NRCM 0240 0FFiCIAL RECORD COPY usa m m i-m o
DECD NUCLEAR ENERGY T
f D
~ '-
i ATIENCY 1
i' 3 8, bd. 5uchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Te!Ix 630668 AEN/NEA RESTRICTED NO. IRS DIFFUSION RESTREINTE Title - Titre
~
Spill of Contaminated Water and Contamination of Personnel in the Auxiliary Building Country - Pays Date of Incident - Date de l 'in ci den t February 1, 1982 United States Type of Reactor - Type de r6acteur t
l PWR l
Plant - Centrale Licensee - Ddtenteur du permis d' exploitation
+
Salem Public Service Electric and Gas Company Manufacturer - Fabricant I
Unit N*
- Tranche n,
1 Westinghouse Power -' Puissance First Commercial Operation - gj77 1090 MWe(net)
Date de mise en service Systems or Components Affected - Systbmes ou composants affect 6s Spent Fuel Pool Auxiliary Building Floor (84-foot level)
Initial Plant Condition - Etat initial de la tranche Mode 6 - Reactor Power 0% - Unit Load 0 MWE g
l Wa. y in which Incident was Detected ?
Comment 1.' i n ci d en t a-t-il 6td ddtectd ?
Noticeable Water Leak Radiation Exposure or Racioactivity Belease -
Exposition aux rayonnements ou libsration de radioactivits l
l Non-measurable amounts to at least 12 workers i
Date of Receipt - Date de tsception Date of Distribution - Date de distribution l
Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetses et
{
cnseignements tir6s (signification de l' incident pour la sGret6) doivent figurer
(*
Event Description In order to accomplish a complete shutdown of the Unit 1 component cooling system, the spent fuel pool cooling system had been cross-connected to the Unit 2 spent fuel heat exchangers via an eight inch diameter hard rubber hose.
After being in service a, proximately two days, a leak was discovered on the 84-foot elevation of n
i the Unit 1 auxiliary building.
About three inches of water covered the 64-foot elevation while approximately an inch of water covered the switchgear room floor.
Surf ace contamination levels to 3000 cpm were measured.
No-airborne activity was found in the auxiliary building and no increase was noted on either plant vent monitor. At least 12 persons received contamination to their shoes and trousers.
The spent fuel pool level remained above the minimum 23 feet above the stored fuel.
Cause The temporary hose leaked at the hose connection due to an improper installation of the clamp.
Reason for Reporting to IRS This event was not a significant occurrence since the leak was noticed in time, but it could have been a severe problem if it was not detected until later.
Therefore, this event is reportable pursuant to criterion 6, " Incidents of Potential
]
Safety Significance."
Action Taken The low level radioactive water was contained by the Unit 1 auxiliary building and recovered by the plant radioactive liquid waste system.
The workmen were decontaminated and returned to work. They received no measurable radiation from the incident. The contaminated floor was properly decon-taminated.
The temporary hose was disconnected and later reconnected using double hose 7
clamps meeting the calculated torque specifications.
I i
i l
t i
l e
n n-
---...e, 1
-r
-,.,.e A
v' DECD NUCLEAR ENERGY
{
T g
D AGENCY
{
38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telrx 630668 AEN/NEA RESTRICTED N2.lRS DIFFUSION RESTREIf4TE Title - Titre Failure of Control Rod Guide Tube Support Pins Date de l' incident Countty - Pays Date of Incident June 9, 1982 United States Type of Reactor - Type de r6acteur PWR Plant - Centrale Licensee - D4tenteur du permis d' exploitation tJorth Anna Virginia Flectric and Pnwor Enmnanv
' " ~
Unit N*
- Tranche n*
Westinghouse Power - Puissance First Commercial Operation - 6l78 865 nwe(net)
Date de mise en service Systems or Components Affected - Systbmes ou composants affectds Control Rods, Steam Generator Initial Plant Condition - Etat initial de la tranche 100%
Way in which Incident was Detected ?
Comment l' incident a-t-il 6td ddtect6 ?
l Loose Parts lionitoring System l
Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou lib 6 ration de radioactivit6 l
tione l
Date of Receipt - Date de r6ception Date of Distribution - Date.de distribution Event description, possibic causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
' Description de l' incident, causes possibles, mesures prises ou projetses et enselenements tir6s (signification de l' incident pour la sGrets) doivent figurer m
1
~~
Event Description On May 17, 1982 a refueling outage was begun one week early due to detection by the loose parts monitoring system of a possible foreign object in steam generator "A."
On June 22, 1982, the licensee informed the NRC that photographs of the damaged tube ends below the tube sheet on the hot leg side of steam generators "A" and "C" showed that the tube sheet and tube ends had sustained peening damage over 75% of the tube sheet area.
By July 1, 1982 Westinghouse had completed evaluations of the foreign objects found in the steam generators and identified them as inconel nuts from control rod guide tube hold-down pins in the reactor's upper internals.
The support pins are assembled to the bottom plate of the lower guide tube by bolting (see Figure 1).
They align the bottom end of the guide tube into the core plate relative to the fuel assembly; provide lateral support for the guide tube; and permit removal and insertion of the guide tube from the upper internals (see Figures 2 and 3).
Cause of Event The event was caused by breakage of the pins due to stress corrosion cracking and high stresses on the pin. The parts then migrated to the "A" and "C" steam generators.
Reason for Reporting The loose parts which result from the failure of these pins have potential to migrate and damage other parts of the plant.
In addition, if both of these pins are missing off of one guide tube, there is the possibility of misalignment and a stuck control rod.
This event is being reported pursuant to criterion 2.4, " Degradation of Systems Required to Control Criticality"and criterion 4 "Significant Generic Problems" because similar failures have occurred at Japanese and French naclear power I
plants.
Actions Taken l
No specific action has been decided upon.
It should be noted, however, that in North Anna Unit 2 all of the control rod guide tube support pins were replaced prior to its receiving a low power test license.
I-A
,d' f
r,.
1
) fY'f@
- i!!!n m a m k
k 3 E" e. J a;
- a d l
in n
=
li -
. T i
T T
j
,jr d
I,'l
.I l
.I I
.I i
, j'
- til3'iliU.Il.IRIlihill!ill!Ifjig-
- /.
o 9
!UI?.01TIOI?M?1 hf! f i {}
i l h E
2-r
-g i
I l
l t
i i
I i
j 5 [ [S INWJ:'
Gh5rEddrl $ $m. xf L - UPPER CORE b
59 PLATE RCC GUIDE TUBES
. ' 1-[i;].
s,g x
[
a j.
+
+
e
+
+
y y
+
g 9
p 7:{ f N
CORE N.
s m
j F
\\
+
+
4 r
J.
4
- Y h_~~
(
t,%wJ j
n
- M l
N.
l l
l l
I
. n.
.?
i??'?
[k 'MI
\\
I N~'JtnLn v,,-
i T
fi}f 1.'~ L;iee, i
1 W
REACTOR ASSEMBLY FI GlA RE.
1.
4.~
.e j
a l
y h,
[
f -- %.N
//
h, N.
/.
f, y
p A
g m
+
m s
m T
i l
f 5
B 2
l 3
l 3
a i
e g
i 8
3 f
B I
5 t
9 1
8 0
0 e
{'
k h
)
)
[
s l
l N 5
5 i
B l
S I 5 f
S I
3 N
ll/
s l
s.
I j \\l:
UPPER SUPPORT i
f i
E M]l a>
a w
l 2
- a.
m-l l
PLATE
.~
R !
I 1
i T's !
l l
l)
/
R$
s' si ml
.i s i Fu' d 3], 1' i
i i
u s
l
.=
6 a
7 M
M W
~~
M
~4
~ UPPER SUPPORT 1
~
~~
e, g
COLUMN 8
'j l
l
.e s c.,
-.=
.=
~
(
I RCC GUIDE i
, o
.=
==
TUBE l
J.
m m
l t
a b
N l 0 0
[
G 0
0
$eW$$LIdTl p l
' N b.5S?w$.r$YlQ
~
E UPPER CORE
. e a.. --
., o..--,.
my o m...__ ;
x PLATE s
SUPPORT PIN i
l UPPER INTERNALS ASSEMBLY Ft 6 0 e.E A.
/
r I 'l
/
F l blL P,C 3 RCC GUIDE TUBE
+
6
~
h i
UPPER GUIDE g
TUBE
.s 6
..6 i
a NT;4 i g
%a'"
b s
UPPER SUPPORT
- =m PLATE en-w A
+
~
s es w-1
/
\\
F
\\
BOTTOM O5 I
FLANGE i==
LOWER SUPPORT
\\
l GUIDE
.O O-PINS l =m TUBE l ~
/
r
/ /O O
i
.Q i
i O\\
O
/ _;. _ _(e y
c 3
\\
Oi i
O
'fi 1
\\O M
3 NO
/
O
. 'l]
Nb
'.O'UoD
@0,
'N I
( b$d 4 x
UPPER CORE D'
PLATE l
~
T{
,/D NUCLEAR ENERGY A
'o nGENCY i
t-38, bd. Suchet 75016 Pans INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED i
No. IRS DIFFUSION RESTREINTE Title - Titre Loss of Reserve Station Service Transformer Country - Pays Date of Incident - Date de l' incident United States January 18, 1982 Type of Reactor - Type de r5acteur PWR Plant - Centrale Licensee - DEtenteur du permis d' exploitation Surry Virginia Electric and Power Company
~
Unit N*
- Tranche n*
Power - Puissance First Commercial Operation -
i MWe(net)
Date de mise en service S/73 i
composants affect 6s Systems or Components Affected - Systhmes ou Reserve Station Service Transformer Initial Plant Condition - Etat initial de la tranche 90% power Way in which Incident was Detected ?
Comment l' incident a-t-il sts dstects ?
Operational event i
Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou libsration de radioactivits None Date of Receipt - Date de r6ception Date of Distribution - Date de distribution Evelt description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, resures prises ou projetees et enseignements tir6s (signification de l' incident pour la sCret6) doivent figurer
Event Description While operating at 90% power, a fault occurred on the feeder to the "B" reserve station service transformer (RSST).
Isolation of the fault de-energized the "E" transfer bus and the "2H" emergency bus.
Emergency diesel generator (EDG)
No. 2 started on the undervoltage and re-energized the emergency bus.
Vital bus 2-1, which is fed from the "2H" bus via a sola transformer, was momentarily de-energized, producing a spurious signal which created a turbine runback.
The plant was stabilized at 62% power and 350 MWe.
Cause Ice forming at the high level intake traveling screen caused a blockage of the fish flume.
Water backed up in the fish flume to the point where it exited through a blow hole (vent).
Brackish water exiting the blow hole sprayed on the terminating insulators for the feeder cables of the "B" RSST.
This apparently caused one of the phases to flashover to ground and to an adjacent phase.
The flashover generated a pilot wire differential signal thereby initiating the isolation of the "B" RSST.
Reason for Reporting to IRS Since the RSSTs ensure immediate availability of electric power to shutdown the reactor safely, a loss of this system would be a " Loss of an Essential Support System" (Criterion 2.6) and reportable as such.
Action Taken The immediate corrective action taken was to verify that the No. 2 EDG was supplying power to the "2H" emergency bus.
In addition, measures were initiated to ensure that an alternate source would be available, if required, within eight hours.
I The blockage of the fish flume has since been eliminated. An investigation revealed that the underground cables from the switchyard for the two affected phases had failed. The failed cables were replaced and the "B" RSST was returned to service within seven days.
Corrective measures were also taken to ensure any water exiting from the blow hole would be deflected away from the RSSTs.
.OECD NUCLEAR ENERGY T
g D
{
_g
%GENCY '
. 38, bd. Suchn 75016 Paris INCIDENT REPSRTING SYSTEM TeI. 524.96.93 Telex 630668 AEN/NEA RESTRICTED N 2. lR$
DIFFUSION RESTREINTE Title - Titre Spurious Trip of a Generator Lockout Relay Associated With'A Diesel Generator Unit Country - Pays Date of Incident - Date de l' incident United States November 19, 1981 Type of Reactor - Type de r6acteur PWR Plant - Centrale Licensee - Ddtenteur du permis d ' exploitation Arkansas Power and Light Arkansas Nuclear One
" ""' ' "*'* ~
b'A'#"
Unit N*
- Tranche n
- i Babcock and Wilcox Bower - Puissance
.First Co*mmercial Operation -
836 MU*IDet)
D*te de mise
- D S e r i c e 12/74
' Systems or Components Affected - Systhmes ou composants affectis Generator Lockout Relay Diesel Generator Initial Plant Condition - Etat initial de la t,r a n ch e 1
90% Power Way in which Incident was Detected ?
Comment l' incident a-t-il sts dstects ?
l 1'o Operator Observation Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou libdration de radioactivits None Date of Receipt - Date de tsception Date of Distribution - Date de distribution i
Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
1 Description de l' incident, causes possibles, mesures prises ou projetses et enseigne:nents tires (signification de l' incident pour la sGrets) doivent figurer sur les pages suivantes.
"g O
t'
/
Event Description On November 19, 1981 while operating at 90% power, an alarm actuated alerting operation personnel that the generator lockout associated with diesel generator Number 1 had tripped.
Further investigation revealed that the C phase high speed differential relay had activated, causing the generator lockout relay to trip, even though the associated diesel generator was not operating at the time.
Cause of Event A walk through in the area by personnel revealed that the diesel generator excitor cubicle door may have been opened and reclosed with sufficient force to cause spunousoperation of this differential relay (General Electric'Model CF012BIA).
Since this type of relay is electrically held when the circuits are in use, the relay would have performed properly had the generator been energized. However, the potential exists for the relay to lockout should agitation occur before the diesel generator is energized. Although this condition would trip the relays, the relays could be reset with a delay in time to the diesel generator availability.
Reason for Reporting This type of relay fails to meet seismic qualifications while in a de-energized mode, thereby making this occurrence reportable pursuant to criterion 3, "Significant Deficiencies in Design, Construction, Operation or Safety Evaluation."
Actions Taken All relays were immediately reset.
No.1 Diesel Generator was started and proved operable. As a preliminary measure, caution cards were attached to the control cubicle doors warning against forceful operation of the doors. A preliminary engineering evaluation was performed, and as an interim measure to prevent pre-mature operation of the relays,. the high speed differential relaysfor both No.1 and No. 2 Diesel Generators have been disabled. A review of the seismic analysis was performed to verify qualification. A complete engineering evaluation is underway to determine a permanent solution and provide an assessment of relay adequacy.
I l
~
~ 'l-s T
7 0
/2 NUCLEAR ENERGY g
NCY
~ :
od. suchet75016 Paris INCIDENT REPCRTING SYSTEM Q
.524.96.93 Ax 630668 AEN/NEA 1
RESTRICTED DIFFUSION RESTREINTE Nr. IRS Title - Titre Cracked Hydraulic Speed Control Cylinders on Main Steam Isolation Valves I
Date of Incident - Date de l' incident Country - Pays United States June 2. 1982 Type of Reactor - Type de rdacteur i
BWR f
Licensee - Dstenteur du permis d* exploitation I
Plant - Centrale j
Iowa Electric Light and Power Company Duane Arnold
" ' ' "* ~
I Unit N*
- Tranche n' General Electric
_l First Commercial Operation -
Sj74
}'
Power - Puissance 545 nwelnet)
Date de mise en service l
Affected - Syst mes ou composants affectss
)
Systems or Components Hydraulic Speed Control Cylinders, Main Steam Isolation Valves, Reactor Protection System Initial Plant Condition - Etat initial de la tranche 0% power - normal operation surveillance testing Hay in which Incident was Detected ?
Comment 1* incident a-t-il dtd ddtects ?
Surveillance Testing Racioactivity Release -
Radiation Exposure or Exposition aux rayonnements ou libdration de radioactivits None i
Date of Receipt - Date de rdception Date of Distribution - Date de distributidn possible causes, actions taken or planned and lessons learned Ev e'n t description, included in the following pages.
(safety significance of incident) should be causes possibics, mesures prises ou projetses et Description de l' incident,
- c~ nts tirse (sienification de l' incident peur la sGret6) doivent figurer
~
, p-Event Description During normal operation while perfonning surveillance testing on the main steam isolation valve (MSIV) closure instrumentation, the position switch ZS4415 did not provide a closure signal to the reactor protection system within the < 10%
valve closure limit.
As required'by a technical specification, the MSIVs were closed within eight hours.
During the subsequent inspection of the Rockwell MSIVs, valves CV4412 and CV4415 were found to have cracked hydraulic speed control cylinders.
On CV4415 the cap screws holding the speed control valve manifold assembly to the hydraulic cylinder were broken and the as'sembly had fallen into the valve operator. There have been no previous similar occurrences.
Cause The cause of the cracked Sheffer MSIV hydraulic cylinders and failed speed control assembly cap screws was due to a violation of procedures during a May 22, 1982 startup.
The operator failed to follow the operating procedures for the opening sequence of the MSIVs, i.e., the four inboard MSIVs were opened before the four outboard MSIVs.
This action resulted in a pressure transient that lifted the main disc while the MSIVs were being stroked open. The resulting impact load was absorbed by the hydraulic speed control unit which cracked the cylinders on MSIVs CV4412 and CV4415, and broke the speed control valve manifold retaining cap screws on MSIV CV4415. The position switch was operable, but out of adjustment with the probable cause being the pressure transient.
Reason for Reporting to IRS The above circumstances make this occurrence reportable to criteria 2.3, " Loss of Containment Function or Integrity," and 3.0, "Significant Deficiencies in Design, Construction, Operation or Safety Evaluation."
Action Taken Before startup, position switch ZS4415 will be tested and reset.
The four inboard MSIVs CV4412, CV4415, CV4418 and CV4420, will be inspected and leak rate tested.'
The cracked hydraulic cylinders and speed control assembly will be repaired or replaced. The four inboard MSIV actuators will be rebuilt. A stress analysis indicated that the piping between the inboard and outboard MSIVs was not over-stressed.
A stress analysis using the highest postulated forces will be performed to determine if any portion of the valve assembly was overstressed. To prevent recurrence, plant procedures and operating instructions will be reviewed and revised, as necessary, and a program will be established for additional operator training.
m
{
OE,CD NUCLEAR ENERGY T
A D
,gGENCY
.{
~
L
. 3 8, bd. Suchst 75016 Paris INCIDENT REPORTING SYSTEM Te1. 524.96.93 q
Telex 630 668 AEN/NEA RESTRICTED Na.lRS DIFFUSION RESTREINTE Title - Titre Large Unnoticed Leakage of Service Water from the Containment Fan Cooler Coils and Supply Piping Country - Pays Date of Incident - Date de l' incident United States Ortnher 17. 1980 Type of Reactor - Type de rdacteur PWR Plant - Centrale Licensee - Ddtenteur du permis d 'exploita tion Indian Point Consolidated Edison
" ""' "' ~ I b'A' "'
Unit n* - Tranche n*
g Westinghouse P-o we r - Puissance
.First Co'mmercial Operation - 7/74 873 nwe(nat)
Date de mise en service
' Systems or Components Affected - Systbmes ou ~composants affectds Containment Fan Cooler Unit Reactor Vessel Safety-Related Equipment on or tiear Floor of Containment Initial Plant Condition - Etat initial de la t,r a n ch e 1
l 100% power 1
Way in which Incident was Detected ?
s l
Comment l' incident a-t-il dtd ddtectd ?
e Upon containment entry to repair a power range detector Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou libdration de radioactivitd 1
flone Date of Receipt - Date de riception Date of Distribution - Date de distribution i
Eve::t description, possible causes, actions taken or planned and lessons learned i
(safety significance of incident) should be included in the following pages, j
Nscription de l'incid$nt, causes possibles, resures prises ou projetEes et c c.s c igac=c n t s tirss (signification de l ' it.c ide n t pour la screts) doivent figurer
o
/
Event Description On October 17, 1980, while operating at 100% power, a load reduction was initiated in response to a positive axial flux tilt indication.
A nuclear instrument power range channel was subsequently declared inoperable and de-energized. When the power to the nuclear instrument was removed, a turbine runback from 85 to 60% power occurred. A reactor trip then occurred when an operator erroneously reduced turbine load from 60 to 12% power.
Soon afterward the plant was taken critical and a return to power was initiated.
About four hours later, a second reactor trip occurred from three percent power while an instrument technician was checking the failed power range channel.
Within an hour the reactor was again critical and at operating pressure and temperature, but the plant manager ordered a reactor shutdown to repair the failed power ra;.je detector.
To repair the detector, containment entry was necessary.
Upon entry it was discovered that approximately four inches (125,000 gallons) of water had covered the operation floor. The water had also spilled into the reactor vessel cavity, submerging the bottom nine feet of the vessel.
Cause The immediate cause for the accumulation of the water was that the containment fan cooler unit (FCU) had a service water leak and that the two sump pumps were inoperabl@
one due to a blown fuse and the other because of a stuck float.
This same FCU had emitted a hich level alarm on October 14, 1980.
At that time the alarm was evaluated to be erroneous and the alarm channel was declared inoperable since there was no noticeable change in the FCU leaE detection instrumentation after the FCU was isolated.
Subsequent investigations revealed that the FCU could have been leaking since the isolation valves were leaking.
The operators did not notice the water in containment since leak detection was not available in the reactor cavity and the containment sump level indication lacked a high level alarm.
Reason for Reporting to IRS There are three major reasons that make this occurrence reportable pursuant to criterion 3, "Significant Deficiencies in Design, Construction, Operation, or Safety Evaluation." They are as follows:
(1) The licensee lacked affirmative action.with respect to correcting the root cause of the FCU/ service water leaks to prevent the event, (2) There exist inadequate procedures for prevention and/or corrective maintenance l
procedures associated with suspected erroneous FCU instrumentation readings and alarms and (3) The inoperability of both sump pumps to remove water accumulation.
In addition, since the reactor vessel was in contact with cold water while critical, an unusual amount of thermal stress was present.
This represents a problem reportable pursuant to criterion 6, " Incidents of Potential Safety Significance."
e 4
2-Action Taken Tests were done on components which came in contact with the cold brackish water to observe the effects of thermal stress and the possibility of corrosion and incipient cracking. All tests showed that no relevant crack-like indications or corrosion damages were observed.
Conduits and penetrations for the incore instrumentation have been 100%
nondestructively examined by the licensee and verified by an NRC contractor.
relevant adverse conditions were detected.
No The level instrumentation already present in the containment sump has been improved and a high water level alarm was installed.
Level instrumentation was also added to the reactor vessel cavity.
Radiation monitors were installed on the service water return line from the fan cooler units.
..m
.... -.