Information Notice No. 86-102:Repeated Multiple Failures of Steam Generator Hydraulic Snubbers due to Control Valve Sensitivity
SSINS No.: 6835
IN 86-102
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
December 15, 1986
Information Notice No. 86-102: REPEATED MULTIPLE FAILURES OF STEAM
GENERATOR HYDRAULIC SNUBBERS DUE TO
CONTROL VALVE SENSITIVITY
Addressees:
All nuclear power reactor facilities holding an operating license or a
construction permit.
Purpose:
This notice is provided to alert recipients of a potentially significant
safety problem pertaining to recent events in which the steam generator
hydraulic snubbers failed to meet their bleed and lockup specifications at
two consecutive refueling outages. The primary cause appears to be control
valve sensitivity to low hydraulic fluid flow velocity. It is expected that
recipients will review the information for applicability to their
facilities. However, suggestions contained in this notice do not constitute
NRC requirements; therefore, no specific action or written response is
required.
Description of Circumstances:
On January 7, 1986, the Portland General Electric Company reported (Licensee
Event Report (LER) 85-13) multiple snubbers which failed to meet their bleed
and lockup specifications at its Trojan Nuclear Plant. The report, and its
supplement dated April 1, 1986, identified three areas of multiple snubber
failures that were discovered during the 1985 refueling outage that began in
May, 1985. These snubber failures were discovered as a direct result of the
expanded inservice testing program which was instituted in accordance with a
recent change to the plant's technical specifications. The prior inservice
inspection program had not required the testing of these snubbers.
The 16 steam generator hydraulic snubbers at Trojan are 900-Kip Anker-Holth
units. Following the failure of the first 2 steam generator snubbers to meet
their bleed and lockup specifications, the remaining 14 were declared
inoperable because of uncertainty regarding the time required to rebuild the
snubbers following testing. All the snubbers were removed and overhauled.
During the overhaul, the snubber seals were found to be degraded and the
hydraulic fluid was heavily contaminated with seal material and rust.
However, as discussed below, this was not the primary cause of the problem
detected during the subsequent 1986 outage.
8612100135
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IN 86-102
December 15, 1986
Page 2 of 4
An engineering evaluation of the effect of the failed snubbers on the steam
generators was not initiated during the 1985 refueling outage because of the
belief that they would not have restricted normal thermal growth. This
decision was based on the snubber-testing company's judgment that the
foreign material in the hydraulic fluid would not have affected the normal
operation of the snubbers because of the relatively large channels through
which fluid would flow under thermal growth conditions. However, in the case
of a seismic or other severe dynamic event, it was determined that the
snubbers would have activated (i.e. , locked up*) and the foreign material
could have blocked the bleed orifice. Because of the manner in which the
snubber control valves are hydraulically interconnected, it is the
licensee's belief that this would have to occur to all four snubbers on one
steam generator before they would become locked in their current position.**
They would remain in this position until a load reversal allowed flow
through the main valves or possibly cleared the bleed port in at least one
snubber.
The revised technical specifications for testing the snubbers required
testing of each snubber that had failed its test during the previous testing
program. Therefore, the 16 steam generator hydraulic snubbers were again
tested during the refueling outage that began in April, 1986. The results of
this testing indicated 12 failures--4 with excessive drag, 4 with high bleed
rates at faulted load, 2 with no bleed rate at faulted load, 1 with
excessive drag and high bleed rate, and 1 with high bleed in compression and
no bleed in tension. The snubbers with no bleed rate cleared themselves upon
load reversal.
There also was an issue of unusual movements of the pressurizer surge line
that was thought for a while to be related to the snubber problems. This is
discussed in Attachment 1.
Discussion:
As a part of its corrective actions during the 1985 refueling outage, the
licensee had all the steam generator snubbers overhauled. Following
overhaul, the snubbers could not meet their safety analysis acceptance
criteria of a
* Note:Common snubber nomenclature uses the term "lock up" to refer
to (1) that point where the main flow path is closed and all flow
is forced through the bleed orifice and (2) the condition where
all flow is stopped and the snubber becomes a rigid strut. To
eliminate any possibility for confusion between the two meanings,
the term "activated" will be used for the first definition.
** In their safety evaluation report (Steven A. Varga's May 30, 1986,
letter to Bart Withers) the Office of Nuclear Reactor Regulation
staff concluded that "...the likelihood of full thermal lock-up
occurring would require that the various contributing factors
would have to affect three or four of the hydraulic snubbers on a
single steam generator."
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IN 86-102
December 15, 1986
Page 3 of 4
maximum drag force of 1,000 pounds at a minimum displacement rate of 0.025
in./min. This was because the snubber activated each time the velocity
approached 0.025 in./min. Following consultation with the reactor vendor,
the acceptance criteria was revised and the snubbers tested satisfactorily.
Because of the reoccurring snubber failures identified during the 1986
refueling outage, the licensee contracted for a detailed root cause analysis
of the snubber failures. This analysis indicated that the low activation
velocity (0.025 in./min) of the steam generator snubbers caused them to
activate at very low fluid velocity through the main flow port. Once the
snubber had activated, all flow was forced through the bleed port. Because
of its extremely small size, this port acted much like a fine sieve.
Apparently the first particle of foreign material would block the port
causing the snubber to lock up. Thus, although contamination of the
hydraulic fluid was a contributor to the problem, it was not the primary
cause.
Based on this root cause analysis, the licensee decided to continue with its
previously made plan to change out the control valves on the steam generator
snubbers. The new snubber control valves have a much higher activation
velocity (6 to 9 in./min) which is still acceptably small compared with that
expected during any significant seismic event. In addition, the new snubber
control valves incorporate a widely used "self-cleaning" poppet valve design
as opposed to the original spring-ball check valve design. In the new
design, the bleed orifices are grooves on the main poppet valve. In this
way, the bleed orifices tend to be self-cleaning whenever there is flow
through the main poppet valve.
All of the Anker-Holth steam generator snubbers were initially designed with
relatively low activation velocities. Therefore, they are suspected of
having the same type of problems as encountered at Trojan. In addition to
Trojan, three other utilities have modified their steam generator snubbers
so that they have activation velocities in the 6 to 10 in./min range.
However, since the root cause of the problem is the selection of an
extremely low activation velocity, as opposed to a design flaw in the
snubbers themselves, the problem may not be limited to only the facilities
having Anker-Holth snubbers.
Attachment 2 to this information notice describes other multiple snubber
failures found at Trojan during the 1985 refueling outage.
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IN 86-102
December 15, 1986
Page 4 of 4
No specific action or written response is required by this information
notice, If you have any questions about this matter, please contact the
Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: Richard J. Kiessel, IE
(301) 492-8119
Attachments:
1. Pressurizer Surge Line Movements
2. Other Multiple Snubber Failures
3. List of Recently Issued IE Information Notices
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Attachment 1
IN 86-102
December 15, 1986
Page 1 of 1
Pressurizer Surge Line Movements
In what had been (until midway through the 1985-86 fuel cycle) a separate
issue, the licensee had been monitoring the unusual movements of the
pressurizer surge line since 1982. A walk-down of this line at the beginning
of the 1985 refueling outage revealed additional movement had occurred
during the last fuel cycle. A consultant was hired to evaluate and analyze
these movements, and had determined that none of the previously identified
potential causes, whether singly or combined, could have produced the
observed movement. However, when he was advised of the possible problems
with the steam generator snubbers, his worst case analysis (i.e., all
snubbers on one steam generator were locked-up) indicated that locked-up
snubbers could have produced the observed movement. This discovery delayed
the submittal of LER 85-13, which was,being prepared at the time.
Testing associated with the root cause analysis demonstrated that the
snubbers on a particular steam generator would not restrict growth of that
loop unless all four snubbers lock-up because the snubber hydraulic lines
were connected in parallel. In addition, based on the results of the thermal
expansion monitoring program conducted during the startup from the 1986
refueling outage, the licensee has determined that most, if not all, of the
observed movement of the pressurizer surge line is expected due to normal
thermal transients experienced by this line during heatups and cooldowns.
Based on these findings, the licensee further concluded that the most likely
cause of the reactor coolant system thermal restraint was due solely to the
inadequate size of the gaps between system components and associated seismic
or pipe whip restraints.
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Attachment 2
IN 86-102
December 15, 1986
Page 1 of 1
Other Multiple Snubber Failures
In addition to the steam generator hydraulic snubber failures, the Trojan
LER identified two other areas of multiple snubber failures. Although not
the subject of this information notice, they are briefly discussed to assist
in identifying all the safety-related failures discussed in the LER.
1. The first additional area of multiple snubber failures was a 25 percent
overall failure rate of small mechanical snubbers (Pacific Scientific
models PSA-1/4 (36 percent failure rate) and PSA-1/2 (17.6 percent
failure rate)).
2. The second additional area of multiple snubber failures involved the
four main steam line hydraulic snubbers (two 70-Kip and two 130-Kip
Bergen-Paterson units). The snubbers were declared inoperable without
testing upon discovery of the steam generator hydraulic snubber
failures.
Additional discussions of multiple snubber failures can be found in IE
Information Notice 84-67, "Recent Snubber Inservice Testing with High
Failure Rates," LER 84-079 for San Onofre Nuclear Generating Station Unit 2
(dated January 25, 1985, and revised March 12, 1985), and LER 85-027 for San
Onofre Nuclear Generating Station Unit 3 (dated May 16, 1985).
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