Information Notice 1996-45, Potential Common-Mode Post-Accident Failure of Containment Coolers
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001
August 12, 1996
POTENTIAL COMMON-MODE POST-ACCIDENT FAILURE OF
CONTAINMENT COOLERS
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a potential common-mode post-accident failure of
containment coolers. It is expected that recipients will review the
information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.
However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.
Description of Circumstances
On February 13, 1996, the licensee for Diablo Canyon determined that the
containment fan coolers were susceptible to component cooling water flashing
in the cooler unit cooling coils during a design-basis loss-of-coolant
accident (LOCA) with a concurrent loss of offsite power or with a delayed
sequencing of safety-related equipment. Diablo Canyon has five fan coolers
per unit.
The containment fan coolers are designed to remove heat from the
containment atmosphere during normal operation.
The fan coolers are also
automatically initiated engineered safety features, which help maintain
containment integrity by reducing post-accident containment pressure by the
condensation of steam and the removal of heat in conjunction with operation of
the containment spray system. The fan coolers are cooled by component cooling
water under both normal and accident conditions.
During a postulated design-basis LOCA with a concurrent loss of offsite power, the component cooling water pumps and the fan cooler fans would lose power.
The component cooling water flow stops almost immediately, while the fans
coast down over a period of minutes.
The first fan cooler will restart on
slow speed approximately 22 seconds after the loss of offsite power.
The
component cc'ling water pumps will restart 26 to 30 seconds after the
accident.
In this scenario, the high temperature containment atmosphere would
be forced across the cooling coils for several seconds with no component
cooling water flow.
This process may cause the stagnant component cooling
water to boil, thereby creating steam voids in the cooling coils and the
9608090031 I-i 5- V4 At- Iu
9 i moos
PDR~~
~9e
s,`
t-;Xc.,96-l-Zc
<2i
IN 96-45 August 12, 1996 component cooling water piping. When the component cooling water pumps
restart, the pumped liquid would collapse the steam voids and may create
significant hydrodynamic loads (waterhammer). The hydrodynamic loads may
adversely affect the integrity of the fan coolers or the associated component
cooling water piping. This problem may cause the fan cooler pipes to rupture
and could cause a significant loss of component cooling water inventory, which
could threaten cooling to emergency core cooling system components. The
licensee for Diablo Canyon evaluated this potential failure mechanism and has
installed a nitrogen pressurization system on the component cooling water head
tank to increase the margin to boiling.
On July 22, 1996, the licensee for Haddam Neck declared all four of its
containment air recirculation fans inoperable and initiated a plant shutdown
The air recirculation fans are the only credited heat removal system for post- accident containment heat loads, and the associated cooling coils are cooled
by the station service water system.
The fans were declared inoperable after
the licensee review of the consequences of two-phase service water flow
following a LOCA concurrent with a loss of normal power.
The licensee
analysis predicted hydrodynamic loads that exceeded the service water piping
and support structural limits.
Discussion
At many plants, containment fan coolers provide a significant safety function
in reducing containment pressure and removing heat from the containment
building. The postulated failure scenario could cause a common failure of all
the containment coolers, thereby potentially challenging the integrity of the
containment building.
In certain cases, failure of the cooling water system
piping could cause a potential containment bypass flow path and may jeopardize
cooling to other safety-related loads.
An individual plant vulnerability to these postulated failures is dependent on
a number of factors. The coastdown rates of the cooling water flow and the
cooler fans, the operating pressure and pressure decay rate of the cooling
water system, the timing of cooling water pump restart, the containment
temperature profile during the design-basis accident, and other site-specific
parameters have an effect on facility vulnerability to this potential failure
mode.
This failure mode would also be applicable to the design-basis main
steam line break in containment.
Westinghouse issued Nuclear Safety Advisory Letter NSAL-96-003, "Containment
Fan Cooler Operation During a Design Basis Accident," to alert their customers
to this concern. NSAL-96-003 is provided as an attachment to this information
notice.
I --1 IN 96-45 August 12, 1996 This information notice requires no specific action or written response.
If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation project manager.
Brian K. Grimes, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts:
James Tatum, NRR
(301) 415-2805 Internet: jetl@nrc.gov
John Tappert, NRR
(301) 415-1167 Internet: jrt@nrc.gov
Alan Levin, NRR
(301) 415-2890
Internet: ael@nrc.gov
Howard Wong, Region IV
(510) 975-0296 Internet: hjw@nrc.gov
William Raymond, Region I
(203) 267-2571 Internet: wjr~nrc.gov
Attachments:
1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96
2. List of Recently Issued NRC Information Notices
- 4-11,g C-411
'LJ
41-1 k-0 t
-a
Westinghouse
Energy
Systs
Business
Usk
NUCLEAR SAFETY ADVISORY LETTER
TIS [S A NOTIFICATION OF A RECENTLY IDENTIFIED POTENTIAL SAFEY ISSUE PERTANrNG TO BASIC
COMPONENU SUPPLIED BY WESTINGHOUSE
THIS INFORMATION 1S BEING PROVIDED TO YOU SO THAT A
REVIEW OF THIS ISSUE CAN BE CONDUCTED BY YOU TO DETERMINE IF ANY ACTION IS REQUIRED.
P.O. Box 335. Pkubuqgb. PA 152304355
Subject: Containet Fan Cooler Operation During a .Dosipn Basis Accidl?{ umber. NSAL-96-003 Basic Coponent: Coota-eat Fan Coole
Date: 6/20/96 Plaits:
Westinghouse and MHI NSSS Plants
,
n
-
r
y
i *
- S
.
%
5
`,
__.-_
_
_
,
..
_ -
_..
.
.
-
-
-r-
.. -.
.
-
.- _. -.,.
.
Transfer of
ormation Pursuant to 10 CFR 21.21(b)
-
'Advisory Eformaion Pursuant to 10 CFR 21.21(cX2)
.
Ye
Yes O
Yes M
&su
5-
.Refcreace:
- -UMM.AR
The purposc of this loe is to provide Westintouse NSSS Ownens information regarding the potential
paformace of some containment fan cooler units (CFCUs) during accidenit conditions which may
compromise the containment cooling system capability.
This Westinghouse Nuclear Safety Advisory Letter deals with the potential susceptibility for the cooling
water of cain CFCUs to flash to steam during a desin basis Lou of Coolant Accident (LOCA) with
ether a cocurrent LOSS of Offsite Power (LOOP) or a delayed sequencing of safety-related equipment.
The applicability of this issue to a given nuclear unit depends on plant specific aspects of the containment
cooling sysem desip and its heat removal system. The Nuclear Regulatory Commission has been
Rotified by _ LER (Reference 1).
Additional iafornation, if required, may be obtained from the originator. Telephone 412-374-5750.
Originator(s):
J. T. Crane
Regulatoiy & Licensing Initiatives
W. R. Ricc. Interim Manager
Regulatory & Licensing Initiatives
AlTACHMM 1 Ul *sWrW Pff
Sheet I of 3
\\
-,A
ISSUE DESCRIPTION
The potenial for this condition wu discovered during an %vestivgeto into Component Cooling Water
(CCW) system design issues by a Westinghouse NSSS Owners The CCW System provides cooling
water directly to the CFCUs in that plant's design. Depending on the specific design conditions, it
may be applicable to plants that utilize cooling waler systems other thin the CCW for their CFCUs
This issue concerns a type of containment cooling deaig that has a single set of CFCUs for heat
removal during both nonmal operating conditions and accident conditis. In the Licensee's design the
CFCUs arc equipped with a swo-speed blower that operates at about 1200 rpm for normal operation
and about 600 rpm for accident conditions. Cooling water to the heat exchanger coils for both normal
and accident operation is supplied by the CCW sytem. The License's containmcn analysis ausumas
the fan coolers operate during postulated accidents, removing beat so as to maintain containment
pressure within design bais limits.
During a postulated LOOP coincident with a design basis LOCA, power is lost to bpth the CCW
pumps and the CFCU blowan. The CCW pumps aem calculated to coast down to 0 rpm within 1-2 seconds while the CFCUI blowers ar calculated to coast down firm a aomiasl speed of 1200 rpm to
600 rpm in approximately 30 seconds and then to 0 rpm an estimated 700 second laser. Accountiig
for diesel-generator start time'seand emergency buss loading sequence, the CFCU blowers are rc-
- energized approximiaeij 20 seconds afler the initiation of the LOOP, and the CCW pumps at re- energized about 30 seconds after the initiation of the LOOP. This timing of events provides for hot, steam-laden cont ent ai to be drawn over the heat exchanger coils at a elatively high velocity for
30 seconds before cooling liquid flow is reestablished to the coils.
The high heat content of containment atmosph=r under accident conditions being drawn over the
CFCU heat exchanger coil with no pumped liquid flow kw been calculated to resit in steaming of the
stationary liquid inventory of the coils. When the CCW pups are re-nergizod, the pumped liquid
flow acts to collapse the steam void in Oe CFCU piping. Collapsing &a steam void is predicted to
mrIlt iG. a wzasAsmer with nfficient energy and force that may impact the integrity of either the
beat exchanger coil or its associated piping, resulting in a lost of inteWy of the CCWs rystm.
To prevent flashing of the CFCU cooling woter, the Licensee has instafled a nitrogen pressurization
system on the CCW system.
TECHNICAL EVALUATION
The following discussion provides several items to consider in evaluating your CFCUs.
In general a plant
u has CFCUs with a 2 speed blower motor and a diesel loading sequence that
results in water flow through the cooler coil later than 15 seconds after a design basis large break
LOCA, may have a similar situation. However, there arn other plant specific aspects that can impact
CFCU functioning. The containment cooling design in this issue has a singk set of fan coolens that
are used for containment cooling during both normal and accident conditions. Flashing and
subsequent void collapse may not apply if a plant has separate sets of equipment where the accident
CFCUs arc not operating prior to a design basis accident. These accident related fan coolers would be
at containment ambient conditions with no forced convective
et transfer at the beginning of the
13i7AMWFWIN
Sheet 2 of 3
I
accident. Since their fmns would not be operating u the beginning of the Postulated accident (as
compated to a single set f fa cooles which would be coasting down) the heat release from a design
basis accacent wouia be ttisfurred to the coil cooling water ju a much slower rate.
Flashing of the CFCU cooling water may not occur in a specific plant design because of factors such
as hea transfer capability, saturation pressure of the cooling systcm. coutdown time of the fan
blower, and coutdown time of the cooling water pump(s). Also, the timing of when ECCS eQuipment
will start in a design basis accident can determine if the CFCUs do or do not have stagnant cooling
water.
In summary, the issue is wbether safety related CFCUs will function as intended during a design basis
event, if they am postulated to experience flshing of the water in the cooling coils. If Rubing can be
postulated, then the issue of whether the CFCUs will maiuntan
euctrl integrity during subsequent
void collapse should be addrssod.
ASSESSMENT OF SAFETY SIGNMCANCE
.
Westinghouse canot deteaine the potential for flashing in the CFCUr on a generic basis because of
the various plant specific designs for the containment cooling system and situation specific heat
umnsfer conditions. Thus &* mott effective action that can be taken is a notification to all
Westinghouse plants.
REEORTABILrY
The NRC is awarm of this imue via an LER (Reference 1).
RECOMENDEDACTINS
Licensees should review d* containment cooling system to determine if their safety related
contaL-=e-t fan ccolerv am nsceptible to cooling water void formation and subsequent void collapse
and waterhammer during a design bais accident.
REFERENCE
1.
Pacific Gas A Etcic letter. PO&E DCL-96-097, April 26, 1996 to U. S. Nuclear Regulatory
Commission, 'Licec
Evet Report 1.96-005-00 (Voluntary) Potential for Flushing in
Containment Fin Cooer Units
Sheet 3 of 3
I
Attachment
August 12, 1996 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
%_
s41TOr.ASDPAP
A
c
-c rssq
.UeiUFU
Urn
raMLAr
Z/U?/vb
WTA_ n aa,
.__^__
96-43
96-42
96-41
96-40
96-09, Supp. 1
96-39
96-38
96-37 irip breaker trom Lracking
of Phenolic Material in
secondary contact assembly
Failures of General
Electric Magne-Blast
Circuit Breakers
Unexpected Opening of
Multiple Safety Relief
Valves
Effects of a Decrease in
Feedwater Temperature on
Nuclear Instrumentation
Deficiencies in Material
Dedication and Procurement
Practices and in Audits of
Vendors
Damage in Foreign Steam
Generator Internals
Estimates of Decay Heat
Using ANS 5.1 Decay Heat
Standard May Vary Signi- ficantly
Results of Steam Generator
Tube Examinations
Inaccurate Reactor Water
Level Indikation and Inad- vertent Draindown During
Shutdown
Degradation of Cooling
Water Systems Due to Icing
08/02/96
08/05/96
07/26/96
07/25/96
07/10/96
07/05/96
06/21/96
06/18/96
06/12/96
All holders of Ls or CPs
for nuclear power reactors
All holders of OLs or CPs
for nuclear power reactors
All holders of OLs or CPs
for nuclear power reactors
All holders of OLs or CPs
for pressurized water
reactors
All holders of OLs or CPs
for nuclear power reactors
All holders of OLs or CPs
for pressurized-water
reactors
All holders of OLs or CPs
for nuclear power reactors
All holders of OLs or CPs
for pressurized water
reactors
All pressurized water
reactor facilities holding
an operating license or a
construction permit
All holders of OLs or CPs
for nuclear power reactors
96-36 OL - Operating License
CP - Construction Permit
IN 96-45 August 12, 1996 This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation project manager.
/s/
Brian K. Grimes, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts:
James Tatum, NRR
(301) 415-2805 Internet: jetlnrc.gov
John Tappert, NRR
(301) 415-1167 Internet: jrt~nrc.gov
Alan Levin, NRR
(301) 415-2890
Internet: ael nrc.gov
Howard Wong, Region IV
(510) 975-0296 Internet: hjw~nrc.gov
William Raymond, Region I
(203) 267-2571 Internet: wjr~nrc.gov
Attachments:
1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96
2. List of Recently Issued NRC Information Notices
^
.
An,%
...........
t
P^
P.
^
,
...
..
..-..
,
UUIUMM NAME:
U:\\JdK\\LUN1LLK.IN
OFFICE
Contacts*
C/SPLB:DSSA
C/PECB:DRPM
D/DRPM
ALevin
LMarsh*
AChaffee*
BGri
HWong
JTatum
JTappert
WRaymond
DATE
08/08/96
08/08/96
08/08/96
Official Record Copy
s d ;
IN 96-XX
August XX, 1996 This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation project manager.
Brian K. Grimes, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts:
James Tatum, NRR
(301) 415-2805 Internet: jetl nrc.gov
John Tappert, NRR
(301) 415-1167 Internet: jrt@nrc.gov
Alan Levin, NRR
(301) 415-2890
Internet: ael nrc.gov
Howard Wong, Region IV
(510) 975-0296 Internet: hjw~nrc.gov
William Raymond, Region I
(203) 267-2571 Internet: wjr~nrc.gov
Attachments:
1. Nuclear Safety Advisory Letter NSAL-96-003, dated 06/20/96
2. List of Recently Issued NRC Information Notices
DOCUMENT NAME:
G:\\JRT\\CONTCLR.IN
OFFICE
Contacts
C/SPLB:DSSA
C/PECB:DRPM
D/DRPM
Km
ALevin
LMarsh-'i
AChaffep
BGrimes
JTatdm'
JTappertJf 7 A
WRa ondM
U
_
_U
DATE
I
/ t/96
/
96
/ t'/96
/
/96 arts
j
-
-
...
If.
uTTIcial Kecora Gopy
IN 96-XX
August XX, 1996 This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation project manager.
Brian K. Grimes, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts:
James Tatum, NRR
(301) 415-2805 Internet: Jetlnrc.gov
John Tappert, NRR
(301) 415-1167 Internet: Jrt@nrc.gov
Alan Levin, NRR
(301) 415-2890
Internet: ael~nrc.gov
Howard Wong, Region IV
(510) 975-0296 Internet: hjw~nrc.gov
Attachment:
DOCUMENT NAME
List OT' Recently Issued NRC Information Notices
E: :\\JRT\\4 NCR.IN
OFFIM
Contacts
C/SPLB:DSSA
C/PECB:DRPM
D/DRPM
K"E
ALevin 0
+
LMars
AChaffee
BGrimes
HWong
p
r>
JTatum
C
V-I
JTappe
14r
rAx
DATE
I
/
/96
4/196
/ /96
/ /96 Official Record Copy