ML20134H147
ML20134H147 | |
Person / Time | |
---|---|
Site: | Fermi |
Issue date: | 02/03/1997 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
To: | |
Shared Package | |
ML20134H131 | List: |
References | |
50-341-96-17, NUDOCS 9702110176 | |
Download: ML20134H147 (22) | |
See also: IR 05000341/1996017
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U.S. NUCLEAR REGULATORY COMISSION
REGION 3
Docket No: 50-341
License No: NPF-43 .
Report No: 50-341/96017
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Licensee: Detroit Edison Company (Deco)
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Facility: Enrico Fermi, Unit 2 L
Location: 6400 N. Dixie Hwy.
Newport, MI 48166 i
Dates: October 4, 1996 through December 6, 1996
Inspectors: A. Vegel, Senior Resident Inspector
C. O'Keefe, Resident Inspector
A. Kugler, Fermi 2 Project Manager, NRR
Approved by: Michael J. Jordan, Chief, Branch 5
Division of Reactor Projects
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9702110176 970203
PDR ADOCK 05000341
G PDR
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EXECUTIVE SUMARY
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Enrico Fermi, Unit 2
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NRC Inspection Report 50-341/96017
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e On October 4,1996, cross-tie valves to the Residual- Heat Removal (RHR)
l reservoirs were rendered inoperable. Operators failed to recognize that
the condition exceeded a Technical Specification (TS) Limiting Condition
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of Operation (LCO). Once the condition was recognized, and actions to
cross-tie the reservoirs were taken, operators failed to evaluate plant
indications. Operators inappropriately detemined that the reservoirs
were cross-tied when one valve had failed in the closed position and
level indications reasonably demonstrated that the reservoirs were :
isolated. This rendered the Ultimate Heat Sink (UHS) unavailable, and i
various safety systems inoperable. The plant was in a condition
prohibited by TS for greater than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. This is an apparent i
violation of several cascaded TS. ;
e On October 13, 1996, the Onsite Review Safety Organization (ORS 0)
inappropriately approved a Technical Specification Clarification (TSC)
in an attempt to operate the plant in a condition that was prohibited by
TS rather than requesting a Notice of Enforcement Discretion (NOED) or
an amendment to the TS. This is an apparent violation of TS.
e On November 4, 1996, the plant re-entered operational Mode 5 without
performing' TS required surveillance testing of the Control Rod Block
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Instrumentation. This is an apparent violation of TS.
These apparent violations were due to several significant root causes:
e One of the RHR reservoir cross-tie valves' (Valve F601A) disk separated
from the valve operator. A set screw on the spline was not tuck welded,
as required, to prevent the screw becoming loose and the disc from
disconnecting from the spline.
e Established periodic testing of the knx seservoir cross-tie valves would
not have detected the valve malfunction.
e Operators and work planners failed to recognize the effect de-energizing
bus 72ED had on the UHS. The planners and approving organizations of
the maintenance activity did not recognize that TS LCO had been entered.
e Operators performed an operability evolution of the UHS using non-
seismic instrumentation in lieu of valid safety-related and seismic
instrumentation that they believed was malfunctioning.
e Licensee made a TS interpretation to allow disregarding a valid TS
requirement. This was due to insufficient knowledge of the regulatory
requirements.
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l e Poor communications between maintenance and operations personnel were a
major contributor to missing a TS surveillance. This was compounded by
insufficient knowledge of technical specifications and inadequate
control of work activities on the refueling floor.
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e Inadequate procedure in that all RPV bolting activities were not l
completed prior to declaring a change to Mode 4. i
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Report Details l
The inspectors reviewed several events that occurred during the refueling
outage. The inspectors independently interviewed plant personnel and
evaluated event logs and data.
I. Doerations
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01 Conduct of Operations
01.1 General Comments (71707) ;
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On three occasions during the refueling outage, between October 4 and :
November 4,1996, technical specification requirements were not met. Two
of the three events are discussed below in the OPERATIONS area while the j
third is discussed in the MAINTENANCE area. , ;
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01.2 Residual Heat Removal (RHR) Reservoir Cross-Tie Lines Were Not Opened
Der Technical Soecification (TS) Reauirements
a. Insoection Scone (93702)
The inspectors independently reviewed the various documentation
associated with the October 4,1996 loss of cross-tie capability to the !
RHR reservoirs. The inspectors also interviewed the appropriate i
operations personnel and management. !
b. Observations and Findinas
On October 4, Operations deenergized bus 72ED in preparation for
maintenance. This action removed power to motor operated valves (MOVs)
Ell 50-F602A and F6028. These MOVs are in one of two cross-tie lines for
the Residual Heat Removal (RHR) reservoirs. Technical Specification (TS)
3.7.1.5, Action C, requires that with one cross-tie line for the
Ultimate Heat Sink (UHS) RHR reservoir inoperable, the valves in the
other cross-tie line shall be opened and deenergized within eight hours.
About Eight hours and nine minutes after one cross-tie line was rendered
inoperable, licensed operators realized that this action statement had
not been completed, so the UHS was declared inoperable. The operating
division of shutdown cooling was declared inoperable as a result, which
was reported to the NRC Operations Center per 50.72(b)(2)(iii)(b). The
cross-tie valves in the other division (Ell 50-F601A and F6018) were
promptly opened and deenergized as required. This event was documented
in DER 96-1288.
Shortly after upening valves F601A and F601B to comply with Action C,
control room operators identified that level indications between the two
reservoirs did not agree, as should be expected with open cross-tie
lines. A non-licensed operator was dispatched to compare local (non-
seismic and non-safety related) indications. Each reservoir had two
local level indicators, and the operator determined that three of the
four agreed, with one of the detectors on the Division I reservoir
reading higher. The operation shift was satisfied that the pools were
successfully cross-tied. Operations considered the UHS to be operable.
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About 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> after the original de-energizing of bus 72ED, water was '
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added to the Division I reservoir with no noticeable change in the
Division 2 reservoir water level. The UHS level divergence was
} investigated, and operators determined that the pools were not cross-
tied. Divers were sent into the UHS and determined that the F601A valve
was actually shut while it indicated open. The motor operated actuator
did not cause valve movement. Operators then opened the other cross-tie
valves to comply with TS 3.7.1.5, Action C. Subsequent investigation by ,
the licensee identified a loose set screw in the bull gear on the F601A
actuator. This event was documented in DER 96-1289.
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c. Conclusions i
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The inspectors determined that on October 4,1996, cross-tie valves to i
the Residual Heat Removal (RHR) reservoirs were rendered inoperable.
Operators failed to recognize that the condition exceeded a Technical j
Specification (TS) Limiting Condition of Operation (LCO). Once the !
condition was recognized, and actions to cross-tie the reservoirs were
taken, operators failed to evaluate plant indications. Operators
inappropriately determined that the reservoirs were cross-tied when one
valve had failed in the closed position and level indications reasonably
demonstrated that the reservoirs were isolated. This rendered the
Ultimate Heat Sink (UHS) and various safety systems inoperable.
The inspectors concluded that the plant was in a condition prohibited by
TS for greater than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />. This is an apparent violation of several
cascaded TS.
TS 3.7.1.5 requires the Ultimate Heat Sink, comprised of two one-half
capacity residual heat removal (RHR) reservoirs with the capability of
being cross-connected, shall be OPERABLE with...(g) two reservoir cross-
connect lines, each with two OPERABLE motor operated cross-connect
valves.
e Action (c) of TS 3.7.1.5 requires with one or more reservoir
cross-connect valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> open and de-
energize both valves in at least one cross-connect line and verify
that these valves remain open and de-energized at least once per 7
days. Otherwise, declare both reservoirs inoperable and take the
ACTION of e. below.
e Action (e.2) of TS 3.7.1.5 requires that in OPERATIONAL CONDITIONS
4 or 5, declare RHRSW system, the EESW system and the diesel
9enerator cooling water systems inoperable and take ACTION
required by Specifications 3.7.1.1, 3.7.1.3 and 3.7.1.4.
Cascaded TS 3.7.1.1, ACTION (c) requires that in OPERATIONAL CONDITION 5
with the RHRSW subsystem (s), which is associated with an RHR loop
required by Specification 3.9.11.1 inoperable, declare the associated
RHR system inoperable and take ACTION required by Specification
3.9.11.1.
e TS 3.9.11.1, ACTION requires with no RHR shutdown cooling mode ,
loop OPERABLE, within I hour and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> j
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thereafter, verify the OPERABILITY of at least one alternate l
method capable of decay heat removal. Otherwise, suspend all l
operations involving an increase in the reactor decay heat load l
and establish SECONDARY CONTAllWENT INTEGRITY within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. ;
i Cascaded TS 3.7.1.3, ACTION requires that with one EESW system -
l subsystem inoperable, declare the associated EECW system subsystem
inoperable and take the ACTION required by Specification 3.7.1.2. ;
- e TS 3.7.I.2, ACTION (b) requires in OPERATIONAL CONDITION 4 or 5, ,
determine the OPERABILITY of the safety-related equipment l
associated with an inoperable EECW system subsyst.en and take the i
ACTIONS required by the applicable Specifications.
Cascaded TS 3.7.1.4, ACTION requires with one or more diesel generator ,
cooling water subsystems inoperable, declare the associated diesel ;
generator inoperable and take the ACTION required by Specification l
3.8.1.2.
e TS 3.8.1.2, ACTION (b) requires that with less than the above
required A.C. electrical power sources (One onsite A.C. electrical
power sourco, Division I or Division II, consisting of two
emergency diesel generators] OPERABLE, suspend CORE ALTERATIONS, '
handling of irradiated fuel in the secondary containment,
operations with a potential for draining the reactor vessel and ,
crane operations over the spent fuel storage pool when fuel :
assemblies are stored within. l
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01.3 Failure to Meet TS Reauirements for Control Rods ;
a. Insnection Scone (93702)
The inspectors reviewed documentation associated with a October 13,
1996, On-site Review Safety Organization (ORS 0) approved Technical
Specification Clarification (TSC)96-003. The inspectors also
interviewed various operations personnel and management.
b. Observations and Findinas
During the refueling outage, with several control rods withdrawn in
defueled cells to permit reactor vessel inspections by camera, a problem
was encountered with the refueling bridge. The withdrawn control rods
had blade guides removed to permit room for inspection cameras, and thus
could not be reinserted for lack of support. Reinsta11ation of blade
guides would have required the use of the refueling bridge. However,
the refueling bridge power supply cable shorted and was repaired during
the camera inspections. When the problem with the bridge was repaired,
the refueling' bridge interlock surveillance was required to be performed
before the bridge could be declared operable. This required briefly
placing the mode switch in Startup to verify interlocks functioned.
However, footnotes in Technical Specification Table 1.2 and Technical
Specification Surveillance Requirement 4.9.1.1 to Technical
i Specification 3.9.1, required that all control rods be fully inserted
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prior to placing the mode switch in a position other than Shutdown or l
- Refuel for surveillance performance. ;
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The inspectors determined that the footnotes were slightly different. i
l The footnote to Table 1.2 stated "the reactor mode switch may be placed i
in Run, startup/ Hot Standby, or Refuel position to test the switch i
- interlock functions and related instrumentation provided that the ;
i control rods are verified to REMAIN FULLY INSERTED [ capitals added for '
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emphasis] by a second licensed operator or other technically qualified !
' member of the unit technical staff." The footnote to the technical ,
specification shrveillance requirement stated "the reactor mode switch ;
may be placed in the Run or Startup/ Hot Standby position to test 1
- interlock functions provided that ALL [ capitals added to emphasize the )
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difference in the footnotes) control rods are verified to REMAIN FULLY l
j INSERTED [ capitals added for em>hasis) by a second licensed operator or l
other technically qualified mem>er of the unit technical staff." The '
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clarification was written to interpret that "all control rods" of
i Specification 3.9.1 and that the term control rods applies only to " core
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cells containing fuel and does not include rods withdraw or removed in
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accordance with 3.9.10.2." This clarification is in agreement with
- Improved - Standard Technical Specifications; however, improved
l specifications are not approved for Fermi.
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i On October 13, ORSO approved Technical Specification Clarification (TSC)
!96-003. Based on the interpretation contained in TSC 96-003, the
, refueling bridge interlock surveillance was performed on October 13,
i resulting in the Mode Switch being unlocked and placed in Run and
Startup with some control rods withdrawn. Fermi did not request an
amendment to their existing technical specifications or a waiver of the
, current requirements.
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i TS 3.9.10.2 requires that any number of control rods and/or control rod
i drive mechanisms may be removed from the core and/or reactor pressure
- vessel provided that at least the following requirements are satisfied
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until all control rods and control rod drive mechanisms are reinstalled
, and all control rods are inserted in the core....(a.) the reactor mode
, switch is OPERABLE and locked in the Shutdown position or in the Refuel
position per Specification 3.9.1, except that the Refuel position "one-
rod-out" interlock may be bypassed, as required, for those control rods
- and/or control rod drive mechanisms to be removed, after the fuel
assemblies have been removed as specified in TS 3.9.10.2 (b through e).
l c. [gnelusions
! The inspectors determined that operators had entered Technical
l Specification 3.9.10.2, to allow withdrawing the control rods for the
- inspections. This TS required that the mode switch remain locked in
i Refuel or Shutdown until all control rods were fully inserted. This was
- in conflicted with the licensee's use of TSC 96-003. At the Residents'
- request, NRR Technical Specification Branch reviewed this icsue, and
- determined that Fermi should have complied with TS 3.9.10.2. The
j appropriate action should have been to request a N0ED or amend their
4 cnrrent technical specifications. This is an apparent violation of TS
l 3.9.10.2.
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02.0 System Description
02.1 Description of the RHR Reservoir
The Fermi Ultimate Heat Sink (VHS) is comprised of two 50 percent
Residual Heat Removal (RHR) reservoirs, which can be interconnected
through either of two cross-tie lines, each containing two ball valves. l
The sink is divided to minimize the impact of a below grade breach of
the reservoir but can be cross-connected to ensure the accident analysis
volume of UHS water is available. The technical specification does not
prohibit operation of the reservoirs either cross-connected or not
cross-connected. Each line has a normally open valve with the other
valve in the line normally closed.
03.0 Seouence of Events
The following sequence of events were determined by the inspectors from
reviews of various parameter chart recording and process computer alarm
recordings.
03.1 RHR Reservoir
11:11 am October 4 Bus 72ED was de-energize rendering RHR reservoir
cross-tie valves Ell 50-F602A and F602B
inoperable. (This prevented cross-connection
through the affected line) (The other cross- ,
connect line has valve F601A closed and F601B l
open)
Operating crew does not recognize that they were
in an 8-hour LCO per T.S. 3.7.1.5, Action C.
7:20 pm October 4 Operating crew recognize that they were in T.S.
LCO. They declare the operating division of
shutdown cooling inoperable. Valve F601A was
directed to be opened (F6018 was already open).
Actions for TS 3.7.1.1, 3.7.1.3, and 3.7.1.4
were also entered. Operations verified that no
core alterations or activities with the
potential to drain the vessel were in progress
or scheduled. TS 3.9.11.1 was the most
limiting. Also, the following systems were
affected; secondary containment, star.dby gas
treatmer.t, control center HVAC, D.C. power
sources, A.C. power sources, and various
electrical power components and systems.
7:41 pm October 4 Valve F601A indicates open in the control room
and operators believe that requirements of TS 3.7.1.5 and the associated cascading TS action
requirements were met. The LC0 Actions were
exited.
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Shortly after opening ,
valve F601A: Operators notice difference in RHR reservoir !
level on control room safety related
indications. A operator was dispatched to
investigate. Local, non-seismic level
indications have 3 of 4 in agreement with each
other. The fourth indicator was out of
calibration since 1993 and could not be ,
calibrated during several attempts since 1993. j
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Operations crew determine that the cross-tie
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line was open and that the UHS was operable. ;
11:14 pm October 4 With six minutes romaining of a required four-
hour notification, NRC was notified via ENS
(Event # 31100) of inoperable shutdown cooling.
This notification was subsequently retracted on
October 5 because the licensee deter. wined that
in addition to the loss of a cress-connect line,
a division of electrical power would also be
needed. Therefore, this was beyond the "alone"
stipulation of the 10 CFR 50.72 criteria. (This
was considered to be valid if valve F601A was
OPEN).
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1:49 pm October 5 Water was added to the Division 1 RHR reservoir,
operators noticed that the level in Division 2
did not change. (Observation of control room
indications). A diver was requested to inspect
the cross-tie line valves.
3:49 pm October 5 Valve F602B (one of the two originally affected
when bus 72ED was de-energized) was manually
opened and valve F601A was closed for inspection
of the reservoir by the diver.
With completion of this action, unbeknownst to !
the operators, the UHS was returned to operable I
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5:00 pm October 5 Operators observed that the indications for the
(about) two reservoirs were equalizing. Division I
reservoir was increasing and the Division II was
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decreasing to an equalization level. Operations
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determined that the cross-connection through
l valves F601A/B (established at 7:41 pm on
l October 4) was not open. It was determined that
j TS 3.7.1.5 and cascading TS 3.9.11.1 was not met
(since TS 3.7.1.1, 3.7.1. 3, and 3.7.1.4 were
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exited on October 4, they were also not met and
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not recognized by the licensee). The plant was
determined to be in a condition prohibited by
TS.
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. 1:59 pm October 6 NRC was updated via ENS. The original 11:14 pm
! on October 4 notification was updated and in ;
effect nullified the retraction. Update does not i
clearly state that T.S. 3.7.1.5 was not met for
the entire period.
LATER Diver determines that valve F601A did not open ]
when operated from the control room (valve '
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position lights indicated open).
October 21 The failure of valve F601A was determined to be
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a loose set screw on the valve operator spline
03.2 Failure tt, Meet TS Reauirements for Control Rods
Sept 27 Plant was shutdown for fifth refueling outage
October ?! First fuel shuffle completed. At stopping
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point, a number of peripheral cells were
defueled, the control rods withdrawn, and the
blade guides removed. In-vessel camera
inspections were begun.
9:50 pm October 11 Refueling bridge blew a main line fuse.
Investigation shows the collector brush assembly
for the power cable takeup reel shorted. Enter
LCO 96-0572
October 13 OSRO approves Technical Specification
Clarification 96-003 to permit retesting
refueling bridge.
1:30 pm October 13 Surveillance 24.623, " Reactor Manual
Control / Reactor Mode Switch / Refueling Platform -
Refueling Interlocks," performed. Mode Switch
in Startup/ Hot Standby for about 47 minutes, in
Run for about 7 minutes. Returned to Refuel and
locked upon completion.
6:30 pm October 13 Exit LCO 96-0572. Refueling bridge declared
4.0 Root Cause and Ma.ior Contributors to the Events ,
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Based on interviews of appropriate personnel, the inspectors determined l
the following root causes and contributors existed during and prior to
the events.
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04.1 RHR Reservoir
The following factors contributed to the event.
e One of the RHR reservoir cross-tie valves (Valve F601A) disk
separated from the valve operator. A set screw on the spline was
- not tacked to prevent loosening and becoming disconnected.
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e Operators and work planners failed to recognize the effect of de- ,
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energizing bus 72ED had on the UHS.
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- Operators did not recognize that the plant was in a condition
i requiring action to meet an LCO.
e Operators performed an operability assessment of the UHS using
, non-seismic instrumentation in lieu of valid safety related and
- seismic instrumentation that they believed was malfunctioning. l
j 04.2 Failure to Meet TS Reauirements for Control Rods
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e Licensee made a TS interpretation to allow disregarding a valid TS
, requirement.
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e Insufficient knowledge of regulatory requirements.
05.0 Safety Sianificance
i 05.1 RHR Reservoir
The consequence of this event was minimal because of the conditions of
l the plant during the event. The plant was in the seventh day of an
refueling outage with little decay heat, no activities in progress that
j could result in draining the vessel, no demand for emergency diesels,
and little heat load on the emergency cooling systems. However, the
- safety significance of this event was moderate to high due to the number i
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and specific systems effected. l
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05.2 Failure to Meet TS Reauirements for Control Rods
I NRR Technical Specification Branch determined that the safety l
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significance of this event was low because the cells with withdrawn
control rods were defueled. This event would not have violated improved
technical specification if improved technical specifications were
applicable to Fermi. However, this event signifies a significant
weakness in using technical clarifications to resolve conflict between
technical specifications without either amending or requesting waiver of
the requirements with a N0ED.
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06.0 Corrective Actions
The following corrective actions were either caserved by the inspectors
or verified through documentation reviews. j
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06.1 RHR Reservoir
The licensee implemented some short term corrective actions. The
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l affected valve (F601A) was repaired. All four cross-tie valves' spline
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bushing set screws were recessed and lock-tighten.
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06.2 [ailure to Meet TS Reauirements for Control Rods
The licensee withdrew the technical specification clarification (TSC
96003) on December 20, 1996. The licensee reviewed other current TSCs
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for similar problems and found none. Currently, the licensee has not
issued a LER or DER documenting this issue.
II. Maintenance i
M1 Conduct of Maintenance
M1.1 Mode Chanae Resultina in Missed TS Surveillance ,
a. Inspection Scone (93702)
The inspectors reviewed various logs and documents associated with the
November 4,1996, event when the plant re-entered Operational Mode 5,
from Mode 4 without performing TS required surveillance. The inspectors
also interviewed both maintenance and operations personnel. The
inspectors also interviewed the appropriate maintenance supervisor.
b. Observations and Findinas
On November 4,1996, the plant re-entered Operational Mode 5, from Mode
4, when a reactor vessel head flange bolt was inadvertently detensioned.
Upon identification that not all reactor vessel head flange bolts were
tensioned, the licensee recognized that they were in Operational Mode 5
and reviewed surveillance requirements. Based on this review, the
i licensee determined that Technical Specification (TS) 4.0.4 requirements
were not met, in that not all surveillances were completed prior to
entry into Operation Condition 5. In this case, the surveillance
requirements for TS 3.3.6, " Control Rod Block Instrumentation," for
Operational Condition 5 were not completed. Because this surveillance
had expired before re-entry into Mode 5, there were less than the
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minimum required operable channels of intermediate range monitors per
trip function.
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c. Conclusions
The inspectors determined that on November 5, 1996, the plant re-entered
operational Mode 5 without performing TS required surveillance testing
of the Control Rod Block Instrumentation.
Technical Specification 4.0.4 requires, in part, that entry into an
Operations Condition shall not be made unless the surveillance
requirements associated with the Limiting Condition for Operations have
been performed. On November 5, 1996, entry was made into Operation 1
Condition 5, without the surveillance requirements for Technical '
Specification 3.3.6, " Control Rod Block Instrumentation," being
performed. This is an apparent violation of TS 4.0.4. !
M3.0 Seouence of Events i
M3.1 Mode Chanae Resultina in Missed TS Surveillance
Initial Conditions: Operational Mode 5
6:02 pm November 4 Head tensioning operations initiated. All 68
head studs installed and hand tightened.
7:27 pm November 4 First pass tensioning (5400 psig) complete.
9:04 pm November 4 Second rass tensioning (7200 psig) complete.
Operations was informed of completion of second
pass. Mode change from Operational Mode 5 to 4
was made. Surveillance for Technical
Specification 3.3.6, " Control Rod Block
Instrumentation," would have been due soon if
the plant remained in Mode 5. With the plant in
Mode 4, the surveillance was no longer required.
9:56 pm November 4 Adjustment pass IAW Procedure 35.710.08 l
initiated.
1:50 am November 5 The Adjustment Pass for final set of four studs
completed.
Later, Maintenance personnel find that stud nut
- 27 was inadvertently loosen enough to move by
hand.
Stud #27 was re-tensioned to 7200 psig.
2:15 am November 5 The Nuclear Shift Supervisor (NSS) was notified
that maintenance personnel found stud # 27
loose.
2:35 am November 5 Refuel Coordinator went to control to fully
brief NSS on situation. NSS recognized that
when stud #27 was inadvertently loosen that the
plant re-entered Mode 5, and TS 4.0.4
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! requirements not met because of an expired
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technical specification required surveillance.
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l All. other studs were subsequently checked, no
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other problems identifled.
M4.0 Root Cause and Ma.ior Contributors
j M4.1 Mode Chanae Resultina in Missed TS Surveillance
- e Poor communications between maintenance personnel and Refuel Floor
Coordinator with operations.
4
e Insufficient knowledge of technical specifications.
- e Inadequate control of work activities on refueling floor.
i e Inadequate procedure in that all RPV bolting activities were not
- completed prior to declaring a change to Mode 4.
j M5.0 Safety Sianificance
! M5.1 Mode Chance Resultina in Missed TS Surveillance
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j The safety consequence: and significance of this event was minimal.
However, the significance of the root cause, failure to recognize the
impact of plant conditions on technical specification requirements was
! high and of importance.
M6.0 Corrective Actions
M6.1 Mode Chance Resultina in Missed TS Surveillance
The licensee will revise the administrative procedure MOP 13, " Refueling
Operations," to define some actions for changing from Mode 5 to 4.
Procedure 35.710.008, " Reactor Vessel Head Detensioning and Tensioning,"
will also be changed to provide thumbrules for adjustments to stud
tension. A caution or note will also be provided that will require
stopping and getting the refueling floor coordinator verification if
more than a turn of adjustment is required. Checks of stud elongation
data will be made between the reactor cavity and.the official record l
before adjustments will be made to ensure the correct adjustments were
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made. Finally, the mode change will be made after all trin passes were ,
completed and stud elongation is within tolerances for all studs. l
However, these changes were not developed before the end of the l
inspection and were not planned to be completed until the end of May,
1997.
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V. Manacement Neetines
j X1 Exit Meeting Summary l
The inspectors presented the inspection results to members of licensee
management at the conclusion of the inspection on December 17, 1996. The
licensee acknowledged the findings presented.
l The inspectors asked the licensee whether any materials examined during the
'
inspection should be considered proprietary. No proprietary information was
identified.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
R. Delong, Superintendent, System Engineering
T. Dong, NSSS, Technical Engineering
P. Fessler, Plant Manager, Operations i
J. Plona, Technical Director :
P. Smith, Director, Nuclear Licensing '
W. O'Connor, Manager, Nuclear Assessment
N. Peterson, Supervisor Compliance
A. Antrassian, Licensing Engineer
J. Moyers, Director Nuclear Quality Assurance
R. Newkirk, Supervisor, Licensing i
R. Eberhardt, Director, Nuclear Training
LIST OF ACRONYMS USED
CCHVAC Control Center Heating Ventilation Air Conditioning
CFR Code of Federal Regulations
DECO Detrnit Edison Company
DER Deviation Event Report ,
EECW Emergency Equipment Cooling Water i
HVAC Heating Ventilation and Air Conditioning
LER Licensee Event Report
M0V Motor Operated Valves
NRC Nuclear Regulatory Commission l
NSS Nuclear Shift Supervisor '
OSR0 Onsite Review Organization
RHRSW Residual Heat Removal Service Water
SOE Sequence of Events
S0P System Operating Procedure
TS Technical Specification
TSC Technical Specification Clarification
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