Information Notice No. 92-71: Revision 1:Unexpected Opening of a Safety/Relief Valve and Complications Involving Suppression Pool Cooling Strainer Blockage

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

                               November 30, 1995

                                           SAFETY/RELIEF VALVE 
                                           AND COMPLICATIONS INVOLVING         
                                           SUPPRESSION POOL COOLING            
                                           STRAINER BLOCKAGE


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this revised
information notice to alert addressees to a recent failure of a safety/relief
valve (SRV) to remain closed during steady-state reactor operation and the
attendant complications involving suppression pool cooling including strainer
blockage.  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.


Information Notice 95-47, "Unexpected Opening of a Safety/Relief Valve and
Complications Involving Suppression Pool Blockage," was issued on 
October 4, 1995.  The notice described a failure of a safety/relief valve to
remain closed, and licensee planned followup actions.  This revision provides
clarifying details on the licensee action plan for monitoring leakage through
these valves (using tailpipe temperature monitoring), makes minor corrections
on event details, adds an additional related generic communication and notes
the NRC staff plans to further evaluate implications of SRV leakage.

Description of Circumstances

On September 11, 1995, the Limerick Unit 1 plant was being operated at 
100 percent power when control room personnel observed alarms and other
indications that one SRV ("M") was open.  Emergency procedures were
implemented.  Attempts to close the valve were unsuccessful and within 
2 minutes a manual reactor scram was initiated.  The main steam isolation
valves were closed to reduce the cooldown rate of the reactor vessel.  The
maximum cooldown rate during the event was 69 �C/hr [156 �F/hr].  Before the
SRV opened, the licensee was running the "A" loop of suppression pool cooling
to remove heat being released into the pool by leaking SRVs.  

9511270084.                                                            IN 95-47, Rev. 1
                                                            November 30, 1995
                                                            Page 2 of 5

The licensee has 2-stage vertical discharge SRVs manufactured by Target Rock
Corporation.  This particular valve design, which is oriented such that
condensate collects on the main stage valve seat, is believed to be the cause
of the continued problems with main stage leakage and is unique to the
Limerick units.  Other licensees use Target Rock valves that have a similar 
2-stage design, but that are configured to discharge horizontally without
condensate collecting on the valve seat.  Even though the licensee had
modified the valve bodies to promote drainage of condensate buildup, these
leaking valves were assumed to still have the same leakage problem through the
main stage valve seat.  

Shortly after the manual scram, and with the SRV still open, the "B" loop of
suppression pool cooling was started.  Operators continued working to close 
the SRV and slow the cooldown of the reactor vessel.  Approximately 30 minutes
later, fluctuating motor current and flow were observed on the Unit 1 "A"
suppression pool cooling loop.  Cavitation was believed to be the cause and
the loop was secured.  

After checking out the pump, the "A" pump was restarted, but at a reduced
flowrate of 8 kL/m [2000 gpm].  No problems were observed so the flow rate was
gradually increased to 32 kL/m [8500 gpm].  No problems were observed so the
licensee continued to operate the pump at a constant flow.  A pressure gauge
located on the pump suction was observed to have a gradually lower reading,
which was believed to be indicative of an increased pressure drop across the
pump suction strainer located in the suppression pool.  After about 30 minutes
of additional operation, the suction pressure remained constant.

The rest of the reactor shutdown was routine and there were no further


Safety Relief Valve:

Shortly after the licensee started up following a refueling outage in March
1994, three SRVs ("F," "M," and "S") were leaking, as determined by tailpipe
temperatures which ranged from 79 �C [175 �F] to 104 �C [220 �F].  These
valves had been refurbished and reset before the restart and they were not
leaking when installed.  However, SRVs "M" and "S" had been stroked during a
3550 kPa [500-psi] automatic depressurization system operability test. 
Reactor operation continued from March 1994 until September 1995, except for
two short mini-outages.  Prior to the recent opening of the "M" SRV, SRVs "D"
and "L" were also observed to be leaking.  Tailpipe temperatures of the five
leaking SRVs were reported as ranging between 102 �C [215 �F] to 141 �C 
[285 �F] and, because of prior experience, the leakage was believed to be past
the main stage valve seat.  Tailpipes at Limerick are uninsulated.  
 .                                                            IN 95-47, Rev. 1
                                                            November 30, 1995
                                                            Page 3 of 5

After the September 11, 1995, shutdown, the leaking SRVs were removed and the
"M" and "S" SRVs sent offsite for inspection to determine the root cause for
the "M" SRV opening.  The "M" SRV was found to have been leaking through the
pilot valve; the other four valves were leaking through the main valve. 
Disassembly of the "M" SRV disclosed that the pilot valve disk was badly
eroded; the nose of the disk had been steam cut 360 degrees around the disk
and had separated from the rest of the disk.  The interior of the disk and the
push rod also showed evidence of erosion.  The pilot valve seat was eroded,
but to a lesser degree.  The reason for the initial leakage is not known.  The
pilot valve seat and disk were fabricated of Stellite 6 and Stellite 6B,
respectively, and were not expected to erode so severely.  

The licensee has replaced the five leaking SRVs and has resumed operation. 
The licensee is investigating techniques for identifying whether the leakage
is through the pilot valve or main stage valve.  Until such a technique is
found, the licensee will assume that all leakage is through the pilot valve. 
As discussed in a letter dated October 6, 1995, the licensee has established a
tailpipe temperature action plan.   When the tail pipe temperature exceeds an
"alert" level of 107 �C [225 �F], the licensee will log temperature more
frequently, project when temperature would be expected to reach 135 �C 
[275 �F], and initiate preparations for an outage to replace the affected
valve(s).  The projection will be based on historical trends and industry
experience.  If the temperature reaches the "action level" of 121 �C [250 �F],
a planned outage will be scheduled to replace the affected SRV before tail
pipe temperature is expected to reach 135 �C [275 �F].

Steam leakage through the pilot valve of about 450 kg/hr [1000 lb/hr] is
estimated to cause a tailpipe temperature of 121 �C [250 �F].  Testing has
demonstrated that this amount of leakage will not cause either the pilot valve
or the main stage valve to open.  Actual experience at Limerick Unit 1 showed
that the "M" SRV operated for more than a year with a tailpipe temperature in
excess of 121 �C [250 �F] before it failed.  The last recorded temperature of
the uninsulated tailpipe before the SRV opened was 141 �C [285 �F]; the
temperature was recorded as 146 �C [295 �F] a week earlier. 

The NRC staff plans to further assess the safety implications of pilot valve
leakage, considering possible effects on SRV valve operability and leakage
detection capabilities.  The staff also plans to examine the efficacy of the
Automatic Depressurization System (ADS) operability test and licensee
practices of routinely operating suppression pool cooling to cope with leaking

Suppression Pool:

Limerick Unit 1 has been in commercial operation since 1986 without having had
the suppression pool cleaned; cleaning was scheduled for the 1996 refueling
outage.  The pool of Unit 2 was cleaned during the 1995 refueling outage..                                                            IN 95-47, Rev. 1
                                                            November 30, 1995
                                                            Page 4 of 5

After a plant cooldown following the blowdown event, a diver was sent into the
Unit 1 suppression pool to observe the condition of the strainers and general
pool cleanliness.  Each strainer is a "T" arrangement with two truncated cones
fabricated from perforated plate; the entire cone surface is covered by a
12x12 316 L stainless steel wire mesh.  The suction strainer in the "A" loop
of suppression pool cooling was found to be covered with a thin "mat" of
material, consisting of fibers and sludge.  The "B" strainer had a similar
covering, but to a lesser extent.  These are the two loops that had been used
for suppression pool cooling necessitated by the leaking SRVs.  The other
strainers in the pool were covered with a dusting of sludge.  Debris was
subsequently brushed off the surface of the strainers, and the suppression
pool floor and water were cleaned by use of a temporary filtration system.  It
is believed that, during operation of the suppression pool cooling system,  
the strainer filtered out fibers that were in the pool water.  The resulting
"mat" of fibers improved the filtering action of the strainers thereby
collecting sludge and other material on the surface of the strainer.  The
licensee believes that the SRV opening increased the rate of accumulation on
the strainer surfaces.  The licensee removed about 635 kg [1400 lb] of debris
from the pool of Unit 1.  A similar amount of material had previously been
removed from the Unit 2 pool.

Analysis showed that the sludge was primarily iron oxides and the fibers were
of a polymeric nature.  The source of the fibers has not been positively
identified, but the licensee has determined that the fibers were not inherent
with the suppression pool.  There was no trace of either fiberglass or
asbestos fibers.

Related Generic Communications

�     NRC Information Notice 95-06:  "Potential Blockage of Safety-Related
      Strainers by Material Brought Inside Containment"     
�     NRC Bulletin 95-02:  "Unexpected Clogging of a Residual Heat Removal
      (RHR) Pump Strainer While Operating in Suppression Pool Cooling Mode" 

�     NRC Information Notice 93-34 and Supplement 1:  "Potential for Loss of
      Emergency Core Cooling Function due to a Combination of Operational and
      Post-LOCA Debris in Containment"

�     NRC Bulletin 93-02 and Supplement 1:  "Debris Plugging of Emergency Core
      Cooling Suction Strainers"

�     NRC Information Notice 92-85:  "Potential Failures of Emergency Core
      Cooling Systems caused by Foreign Material Blockage"

�     NRC INFORMATION NOTICE 92-71:  "Partial Plugging of Suppression Pool
      Strainers at a Foreign BWR"
.                                                        IN 95-47, Rev. 1
                                                        November 30, 1995
                                                        Page 5 of 5

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/'d by D. M. Crutchfield

                                    Dennis M. Crutchfield, Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Robert Elliott, NRR
                     (301) 415-1397

                     Jerry Carter, NRR
                     (301) 415-1153

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