United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 97-31: Failures of Reactor Coolant Pump Thermal Barriers and Check Valves in Foreign Plants

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                         WASHINGTON, D.C.  20555-0001

                                 June 3, 1997

                               BARRIERS AND CHECK VALVES IN FOREIGN PLANTS 


All holders of operating licenses or construction permits for pressurized-
water reactor (PWR) plants.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to reported problems discovered at foreign
pressurized-water reactor (PWR) plants.  Degraded conditions of the reactor
cooling pump (RCP) thermal barriers were found after 10 years of service in
the French 900-MWe PWR plants.  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

The degraded conditions were described in a paper presented during the July
1996 NRC/American Society of Mechanical Engineers (ASME) Symposium on Pump and
Valve Testing (NUREG/CP-0152, July 1996) and subsequently discussed with the
French authorities.  After 10 years of service a sample inspection was con-
ducted to determine the condition of a 900-MWe French PWR RCP.  The inspection
was carried out on RCP No. 2 at the Fessenheim Unit 2 Power Station.  The RCP
had completed 95,000 hours of operation.

During the visual examination, a crack was detected on the outside surface of
the thermal barrier housing.  Further detailed examinations revealed other
cracks inside the housing and on the underside of the thermal barrier flange. 
Additionally, the thermal barrier cooling coil isolation check valves were
found to be jammed open. 

Subsequently, other inspections revealed that cracks existed in almost all of
the 900-MWe plant RCP thermal barrier housings.  Examinations performed on a
1300-MWe power station did not reveal the cracking phenomenon.  About
two-thirds of the check valves in the 9O0-MWe and 1300-MWe plants were found
jammed open.  

9705300075.                                                            IN 97-31
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                                                            Page 2 of 3


The RCP thermal barrier, located directly above the pump impeller, prevents
the hot reactor primary coolant (RPC) from reaching the RCP shaft seals. 
Exposure of the shaft seals directly to hot versus cooled RPC would degrade
the seals.  The thermal barrier is cooled by a cooling coil, which circulates
water supplied from the component cooling water (CCW) system.  The thermal
barrier housing shields the cooling coil from the direct thermal effects of
the hot primary coolant.  Check valves are located in the supply lines to the
thermal barrier cooling coil for the purpose of isolating the CCW system from
the reactor coolant system in the event of a cooling coil rupture.  

The safety concerns were identified and evaluated with regard to the potential
consequences of (1) the formation of loose parts, generated by the thermal
barrier housing, which could damage the pump seals, (2) the jamming of the
pump impeller, (3) the rupture of the cooling coil without the check valve
isolation of the CCW system, and (4) the disintegration of the thermal barrier
On the basis of metallographic examinations, it was concluded that inter-
granular brittle rupture of niobium carbides occurred at the grain boundaries
in some housings as a result of the very low ferrite content of the material
(UNS S34700 SS).  Further, it was concluded that the circumferential cracking
located on the inside cylindrical surface at the interface with the base of
the thermal barrier housing resulted from a fatigue phenomenon.  The thermal
barrier flange required further evaluation and testing to understand the root
cause of the cracking of the surface.  

The cause of the jammed-open check valves was attributed to a layer of
metallic oxide deposits generated in the CCW system carbon steel piping.  The
valves were a lift-type check valve.  Lift-type check valve internals,
particularly in the smaller valves, typically have small clearances and
passages for coolant flow and are highly susceptible to corrosion products

The corrective actions taken for these plants included (1) the installation of
alert alarms and monitoring programs to detect high temperatures in the CCW
system coolant at the thermal barrier outlet, (2) the replacement of the
thermal barrier housings with a configuration- modified unit and a material
change, (3) a check on the condition of thermal barrier flanges in 3 years,
(4) the inservice testing of the lift-type check valves during refueling
outages, and (5) the systematic replacement of the lift check valves with
swing check valves during refueling outages.  
In some recent events, similar failures at domestic plants have been reported
with regard to RCP thermal barrier check valves.  PWR components in safety-
related code class systems such as the RCP check valves located in CCW system
flowpaths are subject to the ASME/OM Code inservice testing requirements. 

On November 16, 1993, with Sequoyah Nuclear Plant Unit 1 in cold shutdown and
Unit 2 at 100-percent power, the Tennessee Valley Authority reported that both
units were outside their design basis as a result of eight inoperable check
valves in the Unit 1 component .                                                            IN 97-31
                                                            June 3, 1997
                                                            Page 3 of 3

cooling system piping, located upstream of the RCP thermal barrier heat
exchangers.  The Unit 1 condition was discovered by radiographic inspections
of the check valves.  The inspections indicated that seven of the eight check
valves were stuck in the open position.  The eighth valve was found to be
improperly assembled.  Subsequent inspections of corresponding valves in Unit
2 indicated that seven valves were stuck in the open position. The cause of
the condition was iron oxide corrosion product buildup wedging between the
valve piston and the bonnet.  The Unit 2 valves were cleaned, reassembled, and
returned to service.  The carbon steel bonnets on the Unit 1 valves were
replaced with stainless steel bonnets.

On October 29, 1991, the Comanche Peak Unit 1 stop check valves (SCVs) in the
CCW lines to the RCP thermal barriers were being tested to satisfy inservice
testing requirements.  During the test, five of the eight SCVs failed to
close.  The valves were subsequently manually exercised, after which they
operated as designed.  On November 6, 1991, two of the failed valves were
inspected.  A small accumulation of corrosion products between the plug and
the stem of the valves, and a slight scaling along the bore, were found. 
Larger accumulations of corrosion products that may have been present were
flushed out when the valves were manually exercised.  The root cause of the
event was attributed to the accumulation of corrosion products and less than
adequate preventive maintenance.  Corrective action included the development
of a preventative maintenance procedure to manually exercise these valves.

Similar valve fouling conditions were indicated during inservice testing at
several other plants over the past 5 years.  

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 (NRR) project manager.

                                            signed by S.H. Weiss for

                                       Marylee M. Slosson, Acting Director
                                       Division of Reactor Program Management
                                       Office of Nuclear Reactor Regulation

Technical contacts:  Francis Grubelich, NRR                 
                     E-mail:  fxg@nrc.gov                   

                     Eric J. Benner, NRR
                     E-mail:  ejg1@nrc.gov
Page Last Reviewed/Updated Tuesday, December 03, 2013