Part 21 Report - 1995-080
ACCESSION #: 9501230060
LICENSEE EVENT REPORT (LER)
FACILITY NAME: Perry Nuclear Power Plant, Unit 1 PAGE: 1 OF 9
DOCKET NUMBER: 05000440
TITLE: Slow Control Rod Scram Insertion Times Result in
Technical Specification Violation
EVENT DATE: 12/12/94 LER #: 94-023-00 REPORT DATE: 01/11/95
OTHER FACILITIES INVOLVED: N/A DOCKET NO: 05000
OPERATING MODE: 1 POWER LEVEL: 80
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR
SECTION:
50.73(a)(2)(i)
LICENSEE CONTACT FOR THIS LER:
NAME: Charles R. Elberfeld, Compliance
Engineer TELEPHONE: (216) 280-5264
COMPONENT FAILURE DESCRIPTION:
CAUSE: X SYSTEM: AA COMPONENT: PSV MANUFACTURER: A610
REPORTABLE NPRDS: Yes
SUPPLEMENTAL REPORT EXPECTED: NO
ABSTRACT:
On December 11, 1994, at 2045, during Technical Specification (TS)
Surveillance testing of control rod scram insertion times, TS Limiting
Condition for Operation (LCO) 3.1.3.2 ACTION c.1 was entered due to the
number of "slow" control rods exceeding twenty percent of a ten percent
sample. Enforcement discretion was requested from the NRC and granted.
On December 12, 1994, at 0845, the TS LCO ACTION was exceeded when the
plant was not placed into HOT SHUTDOWN condition.
The cause of this event is attributed to degradation of the batch 314
Viton seating material located in the disk holder sub-assembly used in
some of the Scram Solenoid Pilot Valves (SSPVs) replaced during the last
refuel outage. The degradation of the seating material causes it to
stick to the seating surface after long periods of solenoid energization,
resulting in a delay in the valve opening after solenoid de-energization.
The delay, in turn, causes the subject control rod to exceed its TS
Control Rod Scram Maximum Insertion Time requirement to notch position
43.
The SSPVs containing batch 314 Viton seating material were replaced.
Perry Nuclear Power Plant, General Electric, and the Automatic Switch
Company are working to resolve the SSPV problem. Augmented Control Rod
Scram Maximum Insertion Time testing will be performed to provide further
confidence in the effectiveness of the corrective actions taken.
END OF ABSTRACT
TEXT PAGE 2 OF 9
I. Introduction
On December 11, 1994, at 2045, during Technical Specification (TS)
Surveillance testing of control rod scram insertion times, TS
Limiting Condition for Operation (LCO) 3.1.3.2 ACTION c.1 was
entered due to the number of "slow" control rods exceeding twenty
percent of a ten percent sample of the control rods. This ACTION
required the plant to be in at least HOT SHUTDOWN within twelve
hours. Enforcement discretion was requested, and granted on
December 11, 1994, at 2230, to allow continued testing to identify
control rods that may have been subject to degraded insertion times
and to provide for timely corrective action for identified rods.
The initial duration of the enforcement discretion extended until
December 16, 1994, at 2230.
On December 12, 1994, at 0845, TS LCO 3.1.3.2 ACTION c.1 was
exceeded when the plant was not placed into HOT SHUTDOWN condition.
This event is being reported pursuant to 10CFR50.73(a)(2)(i)(B) as
operation prohibited by the plant's Technical Specifications.
At the time of the event, the plant was in Operational Condition 1
at 80 percent of rated thermal power. The reactor vessel pressure
was approximately 1000 psig with reactor coolant at saturated
conditions.
II. Background Information
Grand Gulf Nuclear Power Station Experience
Perry Nuclear Power Plant became aware of potential problems with
Scram Solenoid Pilot Valves [PSV] (SSPVs) due to difficulties at the
Grand Gulf Nuclear Power Station (GGNS). Control rod scram time
tests performed at the GGNS in May, 1994 identified that 25 of 191
control rod insertion times to notch position 43 were "slow" (0.320
to 0.390 seconds) and that an additional 8 insertion times failed to
meet the TS limits for operable rods (greater than 0.390 seconds).
GGNS root cause analysis endorsed a material problem with SSPV
seating material located in the disk holder sub-assembly supplied in
pre-assembled top head assemblies. The root cause analysis
indicated that a problem may exist with seating materials (Viton
manufactured post-1991). The failure mechanism is believed to be
associated with degradation of the seating material which causes it
to "stick" to the seating surface. This results in a delay in
venting the air actuators on the inlet and outlet scram valves, and
concurrent "slow" control rod scram time to notch position 43.
Testing performed by General Electric (GE) indicates a relationship
between degradation in SSPV response time and the cumulative
solenoid energization time.
TEXT PAGE 3 OF 9
Perry Nuclear Power Plant Experience
The Perry Nuclear Power Plant (PNPP) SSPVs, Automatic Switch Company
(ASCO) Model Number HVA1768161, were replaced during the most recent
refuel outage in 1994 (RFO-4). Of the 177 SSPVs installed during
the outage, 132 were confirmed to contain Viton batch 314 material
which is known to have been manufactured post-1991. Scram time data
was obtained during startup from RFO-4 on July 27 and August 2 - 4,
1994.
The RFO-4 scram time data indicated that the PNPP control rods met
the TS LCO 3/4.1.3.2 for maximum insertion times. However,
statistical analysis of the RFO-4 scram time data supported a
potential that a failure mechanism could exist which would result in
"slow" actuation of SSPVs. An offset (approximately 0.01 second
increase) was observed in scram time distribution from historical
PNPP scram time data for control rods with certain lots of SSPVs
using batch 314 Viton material. Control rods with non-batch 314
SSPVs exhibited a normal distribution. In addition, a relationship
was observed between the cumulative energization time of the SSPVs
and the scram time in that a statistically significant upward shift
in the data mean values was noted at 195 hours energization time.
Scram time data from RFO-4 was insufficient to project SSPV response
time for longer term solenoid energization. TS Surveillance
Requirement 4.1.3-2.c requires control rod scram time testing at
least once per 120 days of power operation. This is significantly
longer than the maximum energization time (23 days) for which PNPP
test data was available. Because scram time offset appeared to
change with SSPV energization time, additional testing to monitor
this phenomenon, prior to the 120 day surveillance interval, was
determined to be prudent.
The SSPVs had been energized for 49 days on September 24, 1994.
Although not required by TS Surveillance frequency, scram time
testing was performed to determine if degradation was occurring in
the response time of the PNPP SSPVs. An initial test sample of 20
control rods was selected to provide a statistical confidence level
of 95% in SSPV performance. Data from the September 24 and 25, 1994
interim testing as a whole did not indicate degradation in
performance of the suspect (Viton batch 314) SSPVs at 49 days of
continuous energization.
III. Event Description
Control rod scram time testing commenced at 09:50 hours on December
10, 1994, in accordance with TS Surveillance Requirement 4.1.3.2.c,
which requires control rod scram time testing at least once per 120
days of power operation. This testing consisted of a designated
initial test sample of 18 control rods.
TEXT PAGE 4 OF 9
On December 11, 1994, at 2045, testing had been completed for 12 of
the initial test sample control rods, with the following initial
results.
- 8 acceptable rods (scram times to notch position 43 met TS 3.1.3.2
LCO), - 4 "slow" rods (scram times to notch position 43 greater
than TS 3.1-3.2 LCO, but OPERABLE),
- 0 inoperable rods.
Results also showed that rods tested met TS 3.1.3.2 LCO times to
notch positions 29 and 13. Upon identification of each "slow" rod,
testing of adjacent rods began in accordance with TS requirements
concurrent with replacement of the affected SSPV prior to additional
sample testing. Testing of the adjacent rods identified additional
control rods which were "slow" to notch position 43. Subsequent
scram time test results for control rods with replaced SSPVs were
acceptable. Specific test results were documented in Letter
PY-CEI/NRR-1896L, "Request for Enforcement Discretion with Respect
to Control Rod Insertion Times," dated December 12, 1994.
On December 11, 1994, at 2045, the scram time for test sample rod 12
was identified as exceeding the maximum scram insertion time limits
of TS 3.1.3.2. TS LCO 3.1.3.2 ACTION c.1 was entered due to the
number of "slow" control rods exceeding 20% of a 10% sample of the
control rods, thus requiring the plant to be in at least HOT
SHUTDOWN within 12 hours. The plant was therefore required to be
placed in HOT SHUTDOWN by December 12, 1994, at 0845.
Enforcement discretion with respect to TS 3.1.3.2 LCO ACTIONS c.1,
c.3, and c.4 was requested from the NRC and granted on December 11,
1994, at 2330, until December 16, 1994, at 2330, to allow continued
scram time testing to identify control rods that may have been
subject to degraded scram insertion times to notch position 43, and
to provide for timely corrective action for identified rods.
Compensatory measures utilized for the duration of the enforcement
discretion included:
1) implementation of corrective action to replace SSPVs for
all control rods identified as having scram insertion
times to notch position 43 exceeding the scram insertion
time limits of TS LCO 3.1.3.2, and immediate corrective
action implementation for control rods with scram
insertion times identified in excess of TS LCO 3.1.3.2
ACTION a.1,
2) reactor power not to exceed 85% rated thermal power, and
3) administrative reduction of the Maximum Fraction of
Limiting Critical Power Ratio (MFLCPR) acceptance criteria
to 0.990 from 1.00 until repairs were completed on the
rods which had been identified as "slow".
TEXT PAGE 5 OF 9
On December 15, 1994, scram time testing was completed for the
complete core population of 177 control rods with a total of four
rods identified as "INOPERABLE" and 26 rods identified as "slow" per
Technical Specifications. By that time, the four "INOPERABLE"
control rods already had their SSPVs replaced and were restored to
OPERABLE status. Corrective actions had been completed for eight of
the "slow" control rods; however, additional enforcement discretion
time was needed to complete corrective actions on remaining "slow"
control rods to exit the HOT SHUTDOWN requirement of TS 3.1.3.2 LCO
ACTIONs c.1, c.3, and c.4. On December 16, 1994, an extension of
enforcement discretion was requested from the NRC for an additional
seven days ending on December 23, 1994, at 2230 or until compliance
with the Technical Specifications had been achieved, whichever
occurred first. The requested extension and its justification was
documented on Letter PY-CEI/NRR-1897L, "Request for Extension of
Enforcement Discretion with Respect to Control Rod Insertion Times,"
dated December 16, 1994. This extension of enforcement discretion
was granted on December 16, 1994, at 1515.
On December 17, 1994, at 2206, NRC enforcement discretion with
respect to TS 3.1.3.2 Control Rod Scram Maximum Insertion Times
expired when compliance with Technical Specifications was achieved.
IV. Cause of Event
The cause of this event is attributed to the degradation, over time
of solenoid energization, of the Viton batch 314 seating material
located in the disk holder sub-assembly used in some of the Scram
Solenoid Pilot Valves replaced during RFO-4. The degradation of the
seating material causes it to stick to the seating surface resulting
in a delay in the valve opening after solenoid de-energization. The
delay, in turn, causes the subject control rod to exceed its TS
Control Rod Scram Maximum Insertion Time requirement to notch
position 43. GE and ASCO testing as well as PNPP diagnostic testing
support this determination.
V. Safety Analysis
The control rod system is designed to bring the reactor subcritical
at a rate fast enough to prevent the Minimum Critical Power Ratio
(MCPR) from becoming less than 1.07 during the limiting power
transient analyzed in Chapter 15 of the USAR. This USAR analysis
shows that the negative reactivity rates resulting from the scram,
with the average response of control rod drives as given in the
specifications, provide the required protection and MCPR remains
greater than 1.07. The occurrence of scram times longer than those
specified should be viewed as an indication of a systematic problem
with the rod drives.
TEXT PAGE 6 OF 9
The Control Rod Drive Hydraulic [AA] (CRDH) system provides the
hydraulic driving head for insertion, withdrawal, and scramming of
control rods. Each control rod (drive) has a Hydraulic Control Unit
(HCU) which provides the scram function. The CRDH system provides
water at 1720 psig to the HCU. Each HCU contains a scram
accumulator as well as several air operated valves to direct the
water pressure to the control rod drive for insertion, withdrawal,
and scramming of the control rod under various conditions. The
Scram Solenoid Pilot Valve (SSPV) is a dual solenoid operated valve
which, when de-energized, vents air from the actuators of both the
scram inlet and outlet valves, allowing them to open and provide a
flow path for CRDH system and accumulator pressure to scram the
control rod. Each of the SSPV dual solenoids is energized from a
separate Reactor Protection System bus (A or B) and both of the
solenoids must be de-energized for the SSPV to perform its function.
The control rod scram is designed to bring the reactor subcritical
at a rate fast enough to prevent fuel damage. The
accidents/transients that are "scram time sensitive" are those where
the scram time can impact on the Minimum Critical Power Ratio (MCPR)
or the vessel overpressurization limit.
In the current fuel cycle, the limiting event for MCPR is the Load
Reject with No By-Pass valve actuation (LRNBP), and for
pressurization it is the Main Steamline Isolation Valve (MSIV) fast
closure with a neutron flux scram (with no credit for the scram
signal from MSIV position). Loss-of-Coolant Accidents do not set
MCPR limits nor lead to overpressurization concerns, and other
events such as Rod Withdrawal Errors or Rotated Fuel Bundles are not
impacted by scram times. Therefore, the examination of the impact
of a delayed scram initiation concentrated on two areas: a
re-analysis of the limiting scram time sensitive events, and reviews
of the assumptions of the original analyses which serve as the basis
for the Control Rod Maximum Scram Insertion Time Specification.
Re-analyses of the Load Reject with No By-Pass and MSIV Closure
events was performed, which assumed the single failure of the
highest worth rod to insert, and the remainder of the rods received
the Reactor Protection System scram signal with an additional delay
of 70 milliseconds from that assumed in the standard analysis. This
70 millisecond time delay corresponds to the time difference (for
notch position 43) between the standard analysis assumption and the
point at which the rod would be declared "inoperable" by the
Technical Specifications. As an example, the standard analysis
assumption for the time to notch position 43 at 1050 psig is 320
milliseconds (0.32 seconds), and the rod would be declared
inoperable if its scram time exceeded 390 milliseconds (0.39
seconds). If any rod scram times to notch position 43 are measured
during scram time testing at up to 70 milliseconds slower than the
times assumed in the standard analysis, they are treated by
Technical Specifications as "slow" rods, but not inoperable. The
TEXT PAGE 7 OF 9
re-analysis assumption of a 70 millisecond delay simulates that
every trippable rod in the core is held up by its scram solenoid to
the point that they would all be as "slow" as allowed without being
declared "inoperable".
The results of the analyses of these two limiting events identified
that the effects of the scram initiation delay were minimal. For
MCPR, the change in Critical Power Ratio during the LRNBP event (the
"delta CPR") was examined. In the original reload analysis for
Cycle 5 (the current fuel cycle), the delta CPR was 0.15. In the
re-analysis, the delta CPR was 0.16, a change of only 0.01 from the
base case. The MCPR limit for each fuel bundle type in the PNPP
core is determined based on performance of various limiting
transients, and for six of the seven fuel bundle types, the current
MCPR limits are set by the Rotated Bundle Analysis, which is not
scram time sensitive. For the remaining fuel type (the 10 gad rod
GE10 fuel), adding the additional 0.01 delta CPR makes the LRNBP the
most limiting transient for that fuel type. This has been accounted
for by implementing an administrative penalty on the parameter used
during power operation to ensure the MCPR limits are met, i.e., the
Maximum Fraction of Limiting Critical Power Ratio (MFLCPR). The
MFLCPR acceptance criteria was administratively reduced to 0.990
from 1.00 until repairs were completed on the rods which were
identified as "slow". Implementing the penalty in this manner
applied the 0.01 delta CPR universally across the fuel types,
although only the one fuel type was limited by the re-analysis. In
practice this had limited impact, since the core was not currently
being operated near the limits.
The change in peak vessel pressure during the MSIV Closure event was
also examined. In the original reload analysis for Cycle 5, the
peak pressure was calculated to be 1294 psig, well below the TS
2.1.3 Safety Limit of 1325 psig. In the re-analysis, the peak
pressure was calculated to be 1296 psig, also maintaining adequate
margin to the 1325 psig limit.
Reviews were performed of the assumptions in the original analyses
that support TS LCO 3.1.3.2 "Control Rod Maximum Scram Insertion
Times". The purpose of these reviews was to determine the
significance of any rods which might experience scram times slower
than those examined in the re-analysis of the LRNBP event discussed
above; i.e., with times more than 70 milliseconds slower than the
original analysis values. The Technical Specifications direct that
these rods are to be declared inoperable. The analyses that serve
as the basis for the ACTIONs of LCO 3.1.3.2 assume that there are as
many as nine control rods that are inoperable (one (1) "stuck" and
eight (8) inoperable but trippable) in addition to seven (7) more
"slow" rods. The eight inoperable rods were assumed to scram slowly
enough that they do not contribute to meeting the scram reactivity
curve. Although this analysis provides the basis for allowing up to
16 rods to be "slow" (in the above described combination of one
stuck, eight "slow" to the point of being inoperable,
TEXT PAGE 8 OF 9
and seven more "slow"), the current Technical Specification limits
the total number of "slow" (including slow to the point of
inoperability) rods in the core at any one time to seven. These
seven can, therefore, be any combination of "slow" or inoperable
rods, as long as no more than one is a "stuck" rod. This is
reflected by ACTIONs a.1, a.3, b, c.2, and c.4.
Therefore, during the discretionary enforcement period, from a
safety analysis standpoint, there is no significance to rods found
to be "slow", since every rod in the core could have been "slow" and
the safety parameters of concern would have continued to be met. In
addition, the situation where several rods are found to be slow to
the point of inoperability is also bounded, from a safety analysis
standpoint, in that the existing Technical Specification permits up
to seven such rods to exist in the core at any one time. This event
is not considered to be safety significant.
VI. Similar Events
A previous similar event was documented by LER 89-030. On November
25, 1989, the malfunction of two SSPVs due to improper seating
material resulted in a violation of Technical Specifications. The
causes for the November 1989 event were inadequate implementation of
the Nonconformance Control Program and personnel error in the
assessment of test results.
An event that occurred on October 1, 1990, resulted in the discovery
of malfunctioning SSPVs from a single lot installed in the plant
during a refueling outage, and resulted in a 10CFR21 notification on
December 11, 1990.
Another previous similar event was documented by LER 91-018-01. On
October 6, 1991, the plant was shut down in accordance with
Technical Specifications as a result of two adjacent control rods
exceeding maximum insertion scram time limits. The cause of the
slow control rods was SSPV failure. A combination of contaminants
found on the valve disk and seats was believed to have formed an
adhesive which could have bound the valve seat. The suspect SSPVs
were from the same lot remanufactured by ASCO in November 1990. The
49 valves from that suspect lot were removed and steps were taken to
reduce the potential for introduction of contaminants into the
valves.
Because none of the previous similar events involved the use of
post-1991 batch 314 Viton components within the SSPVs, it is not
reasonable to expect that the corrective actions associated with
these events could have precluded the December 1994 event.
TEXT PAGE 9 OF 9
VII. Corrective Actions
The SSPVs containing batch 314 Viton seating material were replaced
by December 30, 1994. The replacement valves used were both new
valves with post-1991 non-batch-314 Viton seating material and
refurbished valves with pre-1991 Viton seating material.
PNPP, GE, and ASCO are working to resolve the SSPV problem.
Augmented Control Rod Scram Maximum Insertion Time testing will be
performed to provide further confidence in the effectiveness of the
corrective actions taken.
Energy Industry Identification System (EIIS) codes are identified in the
text as [XX]
ATTACHMENT TO 9501230060 PAGE 1 OF 1
CENTERIOR
ENERGY
PERRY NUCLEAR POWER PLANT Mail Address: Robert A. Stratman
P.O. BOX 97 VICE PRESIDENT -
10 CENTER ROAD PERRY, OHIO 44081 NUCLEAR
PERRY, OHIO 44081
(216) 259-3737
January 11, 1995
PY-CEI/NRR-1901L
United States Nuclear Regulatory Commission
Document Control Desk
Washington, D.C. 20555
Perry Nuclear Power Plant
Docket No. 50-440
LER 94-023
Gentlemen:
Enclosed is Licensee Event Report 94-023 concerning Slow
Control Rod Insertion Times Result in Technical Specification
Violation.
If you have questions or require additional information, please
contact Mr. James D. Kloosterman, Manager - Regulatory
Affairs at (216) 280-5833.
Very truly yours,
CRE: sc
Enclosure: LER 94-023
cc: NRC Project Manager
NRC Resident Inspector Office
NRC Region III
Operating Companies
Cleveland Electric Illuminating
Toledo Edison
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