ML20132G165
| ML20132G165 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/20/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20132G007 | List: |
| References | |
| 50-298-96-30, NUDOCS 9612260192 | |
| Download: ML20132G165 (13) | |
See also: IR 05000298/1996030
Text
. _ _ _ . _ _ . .
.
..
-
-
- _ - _
. - - . - _ _ . - . _ . _ _ _ _ _ _ _ . . - -
- - - - - _ _ _ .-
j
l
.'
,
!
!
j
0
i
4
ENCLOSURE 2
i
l
>
!
U.S. NUCLEAR REGULATORY COMMISSION
!
REGION IV
,
3
Docket No.:
50-298
!,
License No.:
)
.
- !
Report No.:
50-298/96-30
Licensee:
Nebraska Public Power District
Facility:
Cooper Nuclear Station
l
Location:
P.O. Box 98
Brownville, Nebraska
Dates:
November 4 through December 4,1996
Inspector:
Terrence Reis, Sr. Project Engineer
Approved By:
Elmo Collins, Chief, Reactor Projects Branch C
1
ATTACHMENT:
Supplemental Information
l
'
)
.
H
4
9612260192 961220
ADOCK 05000298
0
_,
1
'
.
\\
,
EXECUTIVE SUMMARY
Cooper Nuclear Station
NRC Inspection Report 50-298/96-30
Operations
Until October 1995, the licensee operated the facility inconsistent with the manner
described in the Updated Safety Analysis Report in that residual heat removal (RHR)
was not required to be available to service the spent fuel pool heat loading when
performing full core offloads.
Enaineerina
The licensee apparently failed to update the Updated Final Safety Analysis following
i
a safety evaluation supporting a 1977 license amendment. Specifically, the license
did not reflect the revised heat loading of the spent fuel pool nor licensing basis
j
regarding maximum heat loading into Section X.5 of the USAR. This item,
i
designated a unresolved in NRC Inspection Report 50-298/96-07 remains
unresolved.
The licensee failed to translate design basis information regarding heat loading of
the spent fuel poolinto operational procedures. Specifically, design assumptions
and requirements indicated that an RHR intertie to the spent fuel pool was required
to service the heat load associated with a full core offload and that the maximum
fuel pool heat load would be associated with a full core discharged to the spent fuel
pool 13 days after shutdown. The licensee did not procedurally control the
1
availabiiity of the RHR system and did not restrict the rate of core offload.
The licensee identified that the reactor core isolation cooling system was not
independent of ac power which caused it to be in violation of the station blackout
rule. In assessing the safety significance of the reactor core isolation cooling design
'
error, the licensee incorrectly credited high pressure coolant injection system
operation for the coping period.
-
-
-
. _ .
_
.
. - .
_ .
.-
9
Details
1. Enaineerina
E2
Engineering Support of Facilities and Equipment
E2.1
Reactor Core isolation Coolina (RCIC) System Deficiencv
,
a.
Inspection Scoce
The scope of the inspection was to assess the significance of the design deficiency
described in Licensee Event Report (LER)94-018 and its supplements and to determine if
the corrective actions specified in the LER were sufficient.
b.
Observations and Findinas
in August 1994, the licensee identified during an extended shutdown that Valve RCIC-
MOV-M014, RCIC turbine trip / throttle valve, was inappropriately powered from an
alternating current (ac) source. The licensee's compliance with the station blackout (SBO)
rule,10 CFR 50.63, " Loss of All Alternating Current Power," takes credit for automatic
operation of Valve RCIC-MOV-M014 in an SBO scenario.
As part of the agency's implementation of the NRC Manual Chapter 0350, " Staff
Guidelines for Restart Approval," process, this design deficiency was identified as a restart
item. In NRC Inspection Report 50-298/94031,it was documented that the deficiency had
been satisfactorily corrected prior to restart.
In the original design of the plant, RCIC was assumed to initiate upon a low vessel level
and automatically shutdown upon a high vessellevel. No credit was taken for reinitiation
of RCIC following the initial actuation and shutdown cycle. Accordingly, the turbine
trip / throttle valve was designed as a manual valve which was normally open and
mechanically tripped closed.
The valve was held open by a mechanicallatching mechanism and closed by spring force
when the latch was disengaged. The valve was equipped with an electromechanical (dc
solenoid powered) trip device that would unlatch or trip the valve closed when specified
physical parameters were exceeded. These included reactor vessel high water level, low
suction pressure, high turbine exhaust, low turbine oil, and overspeed. A manual trip was
i
also provided. As such, the functioning of the valve and the RCIC system were considered
to be independent of ac power.
l
In August 1977, Minor Design Change 74-120 was implemented, which installed a motor
operator on the turbine trip / throttle valve and a control switch in the control room. This
t
motor operator was ac powe,9d and was provided so that an operator did not have to be
'
dispatched to the RCIC room to Sccity reset the valve. The motor operator was
considered an enhancement and a convenience and was not considered as a requirement
to meet its design function. Therefore, even though the motor operator was ac powered,
RCIC was considered to be still capable of performing its safety function as described in
l
_ ~. _ . _ _ _ _ ._
_ _ _ _ _ _ _ _ _ . _ _ _ . _ _ . _ . _ _ .
.-
,
!.
2-
)
the Updated Safety Analysis Report (USAR). This description only required RCIC to be
initiated and be capable of short-term (one cycle) operation independent of ac power.
In 1981, NUREG 0737, item II.K.3.13, required that the RCIC system be modified such
that the system would automatically restart upon a low reactor vessel water level signal
after one cycle. To comply with this requirement the licensee modified the existing AC-
powered motor operator for Valve RCIC-MOV-M014 to automatically reset the turbine
trip / throttle valve, Motor Operator M014 would cycle closed, latch, and open. Minor
Design Change 81-003 was implemented to accomplish this change. After implementation
of the NUREG 0737 requirement, RCIC's safety function was not complete after one cycle
and RCIC was now dependent on AC power to remotely reset Valve RCIC-MOV-MO14.
In 1991, the licensee responded to the SBO rule. As part of the coping requirements for
an SBO, a system capable of operating without ac power was needed to maintain vessel
water level. The licensee designated the RCIC system as being capable of fulfilling this
function.- This was done largely because the cognizant manager responsible for the
licensee's response to the SBO rule failed to adequately research the design basis and
configuration of the RCIC system and relied upon out-of-date information that indicated the
RCIC system was ac independent.
Therefore, as a result of inadequate design and document control changes, the licensee
relied upon incorrect information to respond to an NRC-imposed rule. The RCIC system
was not ac independent and accordingly was incapable of performing its intended safety
function in an SBO scenario. This deficiency existe + l rom the time the licensee was
required to comply with the SBO rule in 1991 untilit was corrected prior to restart of the
facility in March 1995. The dependence of the RCIC system on AC power is a violation of
In reporting this deficiency to the NRC as required by 10 CFR 50.73, the licensee indicated
the safety significance of the event was minimal since both the high pressure core
injection (HPCI) system and local manual reset capability of RCIC would still be available
for vessel level control during the SBO event. However, as a result of.an issue identified
by the NRC inspector, it was determined that the HPCI system would not necessarily be
available for the duration of the coping period.
On May 23,1995, the NRC inspector identified that plant procedural requirements were
inconsistent with assumptions and commitments made to the NRC in a letter regarding
SBO coping strategy dated September 30,1991. The NRC had requested that the licensee
perform a heatup calculation of the HPCI room to assess equipment operability. The
licensee opted not to expend resources on such a calculation and instead committed to -
secure HPCI operation after one cycle of operation. This option was considered acceptable
i
by the NRC.
This regulatory commitment was found not to be properly translated into procedures or
instructions. The action taken as a result of this submittal was revision to CNS Emergency
Procedure 5.2.5.1, " Loss of Off-Site AC Power," dated Februarv 23,1992. The actual
-- - -.
__ - - _ - _ _
.~
F
-3-
procedure direction that was implemented required that operators reduce de loads to a
minimum by shutting down the HPCI system if the RCIC system was maintaining reactor
pressure and level. No clarification was provided restricting HPCI operation to one cycle.
The licensee relied on the information as stated in Procedure 5.2.5.1 in making the
statement in LER 94-018, Revision 1, that the safety significance of the RCIC system not
being AC independent was minimal. It indicated that HPCI would be available. However,
HPCI was not analyzed to operate for longer than one cycle (approximately 10 minutes)in
an SBO event and therefore could not be credited for the entire coping period.
The licensee failed to properly translate a regulatory commitment to secure HPCI operation
after one cycle in an SBO scenario into station procedures. A Severity LevelIV was cited
in NRC Inspection Report 50-298/95-08 for Procedure 5.2.5.1 inadequacy.
10 CFR 50.63, Loss of all alternatina current power, requires that each nuclear power plant
must be able to withstand for a specified duration and recover from an SBO as defined in
10 CFR 50.2. It further states that the reactor core and associated coolant, control, and
protection systems must provide sufficient capacity and capability to ensure that the core
is cooled.
The dependence of the RCIC system on AC power for long-term operation following an
SBO constitutes a violation of 10 CFR 50.63 (298/96030-01).
In addition to correcting the evaluation of safety significance in the LER, the licensee:
Conducted a review of the RCIC system to ensure that other components required
for RCIC to meet its safety function and ac independence did not exist.
Developed an internal procedure, " Regulatory Correspondence Control," to ensure
that the information transmitted to the NRC is complete and accurate.
Held tailgate sessions with engineering personnel to stress the importance of
validating information,
i
These corrective actions were discussed in a public meeting on May 4,1995, and were
'
deemed appropriate,
c.
Conclusions
The licensee identified that the RCIC system was not independent of AC power which
caused it to be in violation of the SBO rule.
i
in addressing the safety significance of its f ailure to comply with the SBO rule, the licensee
l
improperly credited the HPCI system as being available for SBO coping.
.
._
_ _ _
-
.. .
__. _-
._
.
-
_
_ __-
.-
.
'
-4-
4
The licensee's corrective actions were appropriate.
E2.2 Performance of Full Core Offloads Durina Refuelino Outaaes
a.
Inspection Scope
,
The scope of the inspection was to determine if the licensee's practice of routinely
performing full core offloads to the spent fuel pool was consistent with its licensing basis
and regulatory requirements.
b.
Observations and Findinas
In NRC Inspection Report 50-298/96-007,Section 7.3, the staff documented that it
considered the licensee's practice of performing full core offloads during refueling outages
on a routine basis to be inconsistent with the licensing basis of the plant that existed on
4
October 19,1995, when the NRC first questioned this practice. The performance of
routine full core offloads during past refuelings outages without performing 10 CFR 50.59
analyses and the failure to update the USAR following a licensing amendment were left as
unresolved in the inspection report.
,
The observations, findings, and conclusions from NRC Inspection Report 50-298/96-07,
Section 7.3, are restated below:
During a recent evaluation of spent fuel pool decay heat removal and
,
!
refueling practices, the NRC staff reviewed licensing basis documents for
Cooper Nuclear Station. The documents included the USAR and documents
,
associated with Amendment 52 to the Cooper license dated September 29,
-
1978 (rerack amendment). In these documents the inspector found that the
4
routine core offload for these plants was described as being a partial offload.
.
In USAR Section 5.5, the licensee describes two spent fuel assembly
offloads, normal and emergency. The USAR states:
<
Case 1 Normal Heat Load
The normal heat load case consists of one freshly discharged
batch (approximately 160 bundles) in addition to the fuel
"
(approximately 160 bundles) from each of the previous
refueling outages.
Case 2 Emeraency Heat Load
The emergency heat load case which produces maximum heat
load results from the unscheduled discharge of the entire core
I
just prior to a scheduled refueling outage. The freshly
discharged fuel contributes the major portion of the heat load.
__
_
. - , .
_
__
_ . . _
_ =-
_
.__
.
_
_ _ _ _ _ _ _ _ . .
. _ _ -
. _ . _
'
..
.
5-
1
This makes the total heat load quite insensitive to the number
of assemblies cooled for more than a few years in the pool.
The maximum normal heat load is listed in USAR Table X-5-1 as 6.38 x
10e Btu /hr with a fuel pool temperature of 125 F, and the emergency heat
load is listed as 16.6 x 10e Btu /hr with a fuel pool temperature of 150 F.
This table lists the heat exchanger heat removal capacity as 3.19 x
108 Btu /hr for each of the exchangers.
Section 5.6 of the USAR describes use of the RHR system to remove heat
from the spent fuel pool. Specifically it states:
The maximum possible heat load is the decay heat at the full
core load of fuel at the end of the fuel cycle plus the remaining
decay heat of the spent fuel discharged at previous refuelings.
The residual heat removal system is operated in parallel with
the fuel pool cooling and demineralizer system to remove this
heat load . . .
If it appeaG that the pool water temperature will exceed
150 F, d e fuel pool cooling and demineralizer system can be
connected by operator action to the residual heat removal
system. This increases the cooling capacity of the fuel pool
cooling and demineralizer system so that a water temperature
below 150*Fis maintained.
In the rerack amendment request dated July 22,1977, the licensee also
describes the normal and emergency heat loads. The normal heat load is
associated with a partial fuel offload. The emergency heat load is associated
with an " unscheduled discharge of the entire core just prior to a scheduled
refueling outage," as stated in Section 3.3 of the submittal. Based on the
reracked conditions, the heat loads would increase to 7.7 x 108 Btu /hr for
the ncrmal offload, with an assumed cooling time of 7 day: after shutdown,
and to 19.8 x 10' Btu /hr for the emergency offload, with a cooling time of
13 days. The licensee also states in the application that:
For the maximum emergency case, . . . it is expected that the
RHR system will be capable of handling the expanded
maximum heat generation of 19.8 x 105 Btu /hr by either
increasing the flow rate or by allowing the pool temperature to
increase to not more that 160 F (the current maximum is
I
150 F) . . . .
In both cases, the anticipated maximum heat loads can be
i
accommodated without any modifications to the present pool
'
cooling system, with minor changes in design values of the
f
.
_
. .
. - ,
.
.
-
e
6-
maximum allowable pool temperature (no more than 10 F
increase) for relatively short periods (maximum of 8 days).
The inspectors observed that the USAR, through revisions dated July 22,
1994, did not reflect the revised heat loads described in the rerack
amendment application. An October 1995 change to the USAR, performed
in response to NRC concerns, describes the heat removal capacity of the
spent fuel pool cooling and RHR fuel pool in a manner that is consistent with
the information included in the rerack application.
In the NRC staff's safety evaluation for the rerack amendment, the staff
stated that it had focused on the higher fuel pool heat loads that would be
generated by the larger number of fuel to be stored. The NRC staff
commented in the safety evaluation about offload practices:
". . . NPPD INebraska Public Power District, the licensee) '
states that the maximum possible heat load in the spent fuel
pool due to annual refueling will be 7.7 x 108 BTU /hr and that
the heat load due to a full core offload will be 19.8 x
108 BTU /hr."
Further, the NRC staff calculated higher estimated heat loads for the offload
cases based on the additional heat generated by the successive offloads
along with higher potential fuel pool temperatures than that calculated by the
licensee, specifically,9.1 x 108 BTU /hr and 138 F for the normal offload and
21.9 x 108 BTU /hr for the full core offload with temperature still maintained
at 150*F based on the use of the residual heat removal system.
The inspector noted that the definition of normal versus emergency offloads
was clear in the USAR and, further, the use of the term " unscheduled' in the
USAR supports the conclusion that partial offloads were to be the normal or
routine refueling outage practice. The NRC staff considers that the practice
of routinely conducting full core offloads and introducing an off-normal heat
load into the spent fuel pool to be a change to the normal practice defined in
the Cooper Nuclear Station licensing basis. Such a change to the operation
of the facility should have been reviewed pursuant to the requirements of
10 CFR 50.59, including clear discussion of the systems relied on to remove
the spent fuel pool decay heat for the duration of the outage, prior to
implementing the practice of routinely offloading the full core.
The inspector noted that no 10 CFR 50.59 safety evaluation was done
regarding this practice of performing full core offloads on a routine basis until
October of 1995. In October 1995, after significant discussion with the
NRC staffs, the licensee revised the USAR to reflect temperature limits and
heat load during refueling and the systems that are relied on to maintain the
spent fuel pool bulk temperature within those limits.
!
- .-
i
'
9
l
-7-
1
The performance of routine full core offloads during past refueling outages
j
without performing appropriate 10 CFR 50.59 reviews and the untimely
updating of the USAR to reflect information submitted to support the 1977
rerack application is an unresolved item (Update of 298/9514-01).
i
Since NRC Inspection 50-298/96-07 was completed, the NRC has revised NUREG-1600,
q
"NRC Enforcement Policy." This revision provides guidance on the disposition of concerns
arising from departures from the USAR. The guidance was promulgated in Enforcement
j
Guidance Memorandum (EGM)96-005 dated October 21,1996.
J
Based on this guidance, the NRC has not found that performing full core offloads on a
i
preplanned and routine basis constitutes a " defacto change" to the facility as described in
the USAR and has aqi found that a 10 CFR 50.59 violation occurred due to the licensee's
practice of performing full core offloads. However, the NRC has determined that violations
i
of NRC requirements did occur in that:
'
The licensee performed full core offloads without ensuring RHR remained available
to service the heat load. This does represent a " defacto change" to the facility for
j
which a 10 CFR 50.59 evaluation was not performed.
4
The licensee f ailed to translate design basis information regarding heat loading into
]
operational procedures. Specifically, there were no procedural controls in place to
ensure the full core offload was not completed prior to a 13-day cooling period.
!
As stated, the licensee demonstrated in October 1995 that the practice was not
inconsistent with its licensing basis and did not constitute an unreviewed safety question.
However, in performing their investigation, the licensee did identify that there were
!
limitations on the evolution that were not adequately controlled in the licensee's programs
!
and procedures.
The current licensing basis for the heat load capacity of the spent fuel pool was
documented in the NRC staff's Safety Evaluation Report (SER) associated with
Amendment 52 to the operating license. In the licensing submittal, the licensee indicated
that the heat load associated with the full core offload assumed a 13-day (312 hour0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br />)
interval between a reactor shutdown and the time a full core offload was completed. The
staff's SER documented this time interval as part of the licensing basis. The licensee had
no administrative controls in place to ensure that the offload rate was not exceeded.
During Refueling Outage 16, which began in October 1995, the offload rate was exceeded.
Additionally, the staff's SER documented that the licensee would use the RHR system for
cooling the spent fuel pool when a full core offload was performed. The purpose of this
licensing basis requirement was to ensure the temperature of the spent fuel pool did not
exceed the design temperature of 150 F. In October 1995, while in the prc, cess of
performing a full core offload, RHR Train B, which is the only train that can physically
service the fuel pool, was not available. However, the situation was corrected prior to the
licensee discharging more than 1/3 of the core. The SER documents that the normal spent
--- . - . - .
--
- . - . - - - . - . - .
- - - . _ .
. . . -
_ . . -
>
+
-8-
fuel pool cooling system is capable of handling this heat load it is not known if RHR
Train B was available to serve the spent fuel pool during previous outages, but the licensee
acknowledges that it had no_ outage plans or administrative procedures to ensure RHR
Train B remained available.
10 CFR 50.71(e) requires that each person licensed to operate a nuclear power reactor
shall update periodically the final safety analysis report to assure that the information
included in the final safety analysis report contains the latest material developed. The
updated final safety analysis report (USAR) shall be revised to include the effects of all
changes made in the facility or procedures as described in the final safety analysis
report (FSAR) and all safety evaluations performed by the licensee in support of requested
license amendments.
In its July 22,1977, licensing submittal supporting License Amendment 52, the licensee
designated the maximum possible heat load in the spent fuel pool due to an annual
refueling as 7.7E6 BTU /hr and the heat load associated with a full core offload as
19.8E6 BTU /hr.
In its July 22,1977, licensing ' submittal supporting License Amendment 52, the licensee
designated a 13-day cooling time interval between a reactor shutdown and the time a full
core offload is completed.
Nebraska Public Power District, the licensee for Cooper Nuclear Station, failed to revise the
USAR to include the effects of:
1.
A safety evaluation performed by the licensee dated July 22,1977,in support of
License Amendment 52 in that on October 20,1995, USAR Table X-5-1 designated
the maximum normal heat load of the spent fuel pool as 6.38E6 BTU /hr and
emergency heat load (full core offload) as 16.6E6 BTU /hr.
2.
A safety evaluation performed by the licensee dated July 22,1977,in support of
License Amendment 52 in that on October 20,1995, Section 5 of the USAR did not
address that maximum heat loads were associated with a 13-day cooling time
between a reactor shutdown and the time a full core offload is completed.
This aspect of Unresolved item 298/95014-01 to update the USAR to reflect Licensee
Amendment 52 remains open.
10 CFR 50.59(a)(1) requires that a licensee may make changes in the facility as described
in the safety analysis report without prior Commission Approval unless the change involves
a change in the Technical Specifications incorporate in the license or an unreviewed safety
question. 10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of
changes in the facility, to the extent that these changes constitute changes in the facility
as described in the safety analysis, and that these records must include a written safety
'
evaluation which provides the basis for the determination that the change did not involve
an unreviewed safety question.
-
.
_
_-
._.
_
_
.
- _
. - . - . -
_ - - - . . - - . - - - -
_ _ - - - _ - . -
--
>
..
9
On October 20,1995, the licensee's FSAR, Section 8.5.6, stated, in part, that "the
,
!
residual heat removal (RHR) system can be intertied with the Fuel Pool cooling system if
required. This capability increases the spent fuel pool cooling capacity in the event that
such additional capacity is necessitated by removal from the core of an unusually large
.
number of fuel elements. The RHR system - fuel pool cooling system intertie is sized to
remove an emergency heat load . . . from the fuel pool which corresponds to full core off-
loading plus the batch of spent fuel discharged at the previous refueling outage.
In the NRC's safety evaluation supporting License Amendment 52 dated September 29,
1978,it was stated in Section 2.2 that the licensee would use RHR cooling when
performing full core offloads.
On October 20,1995, the licensee changed the facility as described in the safety analysis
l
report in that the facility was not operated as described in the FSAR and a written safety
i
evaluation of the change from the FSAR had not been performed to determine whether this
change involved an unreviewed safety question. Specifically, the licensee was in the
process of performing a full core offload, and the RHR system was not available to assist
the fuel pool cooling system in removing what the FSAR characterized as an emergency
offload. This is a violation of 10 CFR 50.59 (298/96030-02).
Criterion lll of Appendix B to 10 CFR Part 50 requires that regulatory requirements and the
design basis, as defined in 10 CFR 50.2 and as specified in the license application, for
those structures systems and components to which the appendix applies are correctly
translated into specifications, drawings, procedures, and instructions.
In the safety evaluation report which accompanied Amendment 52 to the facility operating
license, the NRC staff acknowledged that the licensee's spent fuel pool and cooling
systems were capable of handling the heat load associated with a full core discharge.
However, this acknowledgement was based on certain design assumptions. In the Safety
Evaluation Report, the staff stated that the maximum fuel pool heatload was associated
with an offload that would occur 13 days after shutdown.
The design basis assumption that the maximum heat load was associated with full core
discharge which was completed in 13 days was not translated into procedures.
Procedure 2.3.2, " Fuel Pool Cooling and Demineralizer System," contained no
administrative controls to ensure that fuel was not loaded at a rate that would exceed the
'
.
13-day assumption. in October 1995, the licensee did exceed this offload rate before the
j
l
full core offload was halted at'1/3 of the core. This is a violation (298/96030-03).
!
I
i-
The two violations (298/96030-02,03) disposition the spent fuel pool operation aspects of
!
Unresolved item 298/95014-01 (Update).
l
c.
Conclusions
The licensee apparently failed to update the USAR, following a safety evaluation
supporting a 1977 license amendment.
4
i
.
. _ . ,
-.-
, . ~
- __
. . - _
_
-
,
,
. . _ _ , _
--
-
1
'
.;-
-10-
'
1
The licensee operated the facility inconsistent with the manner described in the USAR in
i
that RHR was not required to be available to service the spent fuel pool heat loading when
performing full core offloads.
The licensee failed to translate design basis information regarding heat loading of the spent
fuel pool into operational procedures. Specifically, design assumptions and requirements
indicated that an RHR intertie to the spent fuel pool was required to service the heat load
associated with a full core offload and that the maximum fuel pool heat load would be
associated with a full core discharged to the spent fuel pool 13 days after shutdown. The
licensee did not procedurally control the availability of the RHR system and did not restrict
the rate of core offload.
II. Manaaement Meetinas
XI
Exit Meeting Summary
The inspector met with licensee management on November 7,1996, and presented the
findings. At that time, one of the violations involving operation of the spent fuel pool was
characterized as a " defacto change" to the facility for performing full core offloads when
wording in the USAR indicated that the performance of a full core offload was an
j
emergency condition. The licensee objected to that characterization.
On December 4 and 20,1996, the inspector formally exited with the Manager, Nuclear
Safety and Licensing. At this time, the issue discussed above was characterized as
presented in this report, which is an unresolved item.
_ . . _ _ _ _ . . _ _
_
_
. . _ - _ _ . . . . . . _ _ . _ _ _ _ .
.m.._.
. . . _ . . .. ..
_
-. . _ . _ . . . . _ _ _ . _ _ _ _ . _ . .
>
l
l
,*
Sucolemental Information
[
PARTIAL LIST OF PERSONS CONTACTED
!
Licensee
Brad Houston, Nuclear Safety and Licensing Manager
Robert Godley, Manager, Plant Engineering
!
Mike Peckham, Plant Manager
l
Rick Gardner, Operations Manager
Jack Dillich, Maintenance Manager -
1
!
INSPECTION PROCEDURES USED
IP 92903: Followup - Engineering
ITEMS CLOSED
i
RCIC System inoperable in Station Blackout Scenario
'
298/96030-01
Failure to Comply with Station Blackout Rule
Plant Procedural Requirements Inconsistent with Station
Blackout Assumptions
I
ITEMS OPENED AND UPDATED
l
298/96030-01
Failure to Comply with Station Blackout Rule
l
l
298/96030-02
Failure to Comply with 50.59 in that RHR System not Available
298/96030-03
Failure to Translate Design Basis information into operational
Procedures
298/95014-01
(UPDATE) The aspects of this URI regarding the operation of
the spent fuel pool are closed. The apparent failure to update
the USAR remains unresolved.
i
f
,.
-
- - .
-
-
.
.
.-