ML17311A576
| ML17311A576 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 01/09/1995 |
| From: | Gwynn T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17311A574 | List: |
| References | |
| 50-528-94-35, 50-529-94-35, 50-530-94-35, NUDOCS 9501180121 | |
| Download: ML17311A576 (54) | |
See also: IR 05000528/1994035
Text
APPENDIX 8
U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-528/94-35
50-529/94-35
50-530/94-35
Licenses:
NPF-51
Licensee:
Arizona Public Service
Company
P.O.
Box 53999
Phoenix,
Facility Name:
Palo Verde Nuclear Generating Station,
Units
1
~ 2,
and 3
Inspection At:
Wintersberg,
Inspection
Conducted:
November 15-17,
November
29 through
December
2,
and
inoffice review until December
16,
1994
Inspectors:
Dr. Dale A. Powers,
Chief, Maintenance
Branch
Division of Reactor Safety
John
E. Whittemore,
Reactor
Inspector,
Maintenance
Branch
Division of Reactor Safety
Accompanying
Personnel:
Margaret
S. Chatterton,
Nuclear Engineer,
Reactor
Systems
Branch, Office of Nuclear Reactor Regulation
Brian E. Holian, Senior Project Manager,
Project
Directorate
IV-2, Office of Nuclear
Reactor Regulation
Approved:
I5fs-
omas
.
wynn,
erect r.
ivy ion o
eactor
a e y
t
Ins ection
Summar
Areas
Ins ected
Units
1
2
and
3
Routine,
announced,
followup inspection
of Licensee
Event Report 528;
529: 530/94-002-00
and its two supplements,
which discussed
Technical Specification limiting conditions for operation that
were not supported
by safety analysis.
9501180121
950111
ADOCK 05000528
9
'I
'l
I
i
'
,~
Results
Units
1
2
and
3
Plant
0 erations
-2-
Not applicable during this inspection.
Maintenance
Not applicable during this inspection.
~f
Although generally untimely. engineering
decisions
related to revised
core physics limits were technically appropriate.
There were no
technical
inadequacies
found in the licensee's
revised administrative
limits that had supplanted
Technical Specification limits (Section 1.2).
There were missed opportunities to identify and correct deficiencies.
The issue of subcritical control element
assembly
bank withdrawal
reanalysis internally reported
on March 17.
1994.
was originally
recognized
by the licensee's staff in November
1993.
The issue of
moderator temperature coefficient discrepancy
could have
been identified
during the development of previous core reload analyses
and
reviews
by the fuel vendor
and the licensee staffs.
This issue of
Mode 6 boron concentration
had the potential to be discovered
at each of
14 reloads
before it was found (Section 1.2. 1, 1.2.2
~ and 1.2.5).
The licensee's staff did not consider analytical
inputs
and reload
analyses
assumptions
for matters
such
as core azimuthal tilt accident
analyses,
which had led to plant operational limits such
as
3 percent
core azimuthal tilt. to constitute design
bases
(Section 1.2.3).
The licensee's
staff inappropriately derived
an
NRC staff viewpoint from
reviewing other
licensee
correspondence,
rather than pursuing its own
Technical Specification
change.
The licensee's
practice of abandoning
Technical Specification limits for the use of more restrictive
administrative limits to control safety analysis
assumptions
as
a
prolonged or routine practice instead of requesting
NRC review and
revision of Technical Specification limits was not prudent.
The
licensee's
implementation of its reportabi lity process
and its frequent
requests
to the
NRC for time extensions
related to licensee
correspondence,
was
an area in need of significant improvement
(Section 1.2.4).
The licensee's
program for evaluating
vendor technical
information
inappropriately
exempted fuel vendor information (Section 1.3. 1).
1
,
-3-
The licensee's staff had recognized
and was working toward resolving
a
weakness
that there
was not
a governing procedure to ensure that
low-tier procedures,
plant changes,
and concerns
about reload safety
analyses
were captured
for subsequent
considerations
in unit cycle bases
documents
(Section 1.4. 1).
Analytical inputs
and analyses
for the Unit 2 Cycle 5 fuel thermal
performance
were reasonable
and acceptable
(Section 1.4.2).
10 CFR 50.59 safety evaluations
for recent fuel design
changes
were
brief with an almost complete reliance
on the fuel vendor's analytical
results
(Section 1.4.3).
Certain training records
had been lost.
A new training procedure
eliminated reload analysis continuing training.
One group had personnel
qualification cards;
however,
the cards
had not been
used.
There was no
system for tracking or evaluating training.
The present continuing
training program for reload analysis
was not viable (Section 1.4.4).
Plant
Su
ort
The corrective action tracking process
was weak in that it allowed
unnecessary
vulnerability to the failure to track and implement
necessary
corrective actions
(Section 1.3. 1).
There were programmatic
requirements
in place to assure that commitments
implemented within procedures
would not be eliminated
by procedure
revision.
However, there were no safeguards
to assure that database
searches
were actually performed
and correctly interpreted;
therefore,
there
was
a vulnerability to the inadvertent eliminatation of
commitments
(Section 1.3.2).
~
Hang ement Overview
The long-term outstanding
issues
discussed
in Licensee
Event Report 528;
529: 530/94-002-00
and its supplements
revealed historical
management
difficulties in ensuring priority consideration to resolving with NRC
technical
issues potentially impacting safe plant operation
(Section 1.1).
Upon identification of a recent issue involving the exercising of
control element assemblies
in Nodes 3. 4.
and
5 that could violate the
assumptions
in the safety analysis,
a special
plant review board
met in
a timely manner (the
same day) to discuss
the issue.
Since
a proposed
solution had not been finalized, the board took positive action to
require prereview by the board of future control element
assembly
withdrawals (Section 1.2. 1).
J
0
Summar
of Ins ection Findin s:
~
Violation 528;
529; 530/9435-01
was opened
(Section 1.2.1).
~
Licensee
Event Report 528;
529; 530/94-002
and its Supplements
01 and
02
were closed
(Section 1.2.6).
~
Inspection
Followup Item 528;
529; 530/9435-02
was opened
(Section 1.4.4).
Attachment:
~
Attachment
- Persons
Contacted
and Exit meeting
E
I
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-5-
DETAILS
1
ENGINEERING FOLLOMUP
(92903)
1. 1
Backcaround
~
.
Technical Specification 3.3.1.
required three of four core protection
calculators to be operable in Modes
1 and 2.
However, the safety
analysis
and plant operating procedures
required that the core
protection calculator bypass
be operable during any subcritical
operation with the reactor trip breakers
closed.
Technical Specification 3.9. 1. required either
a K-effective of less
than or equal to 0.95 or a boron concentration of greater
than or equal
to 2150
ppm, whichever was more restrictive,
when in Mode 6.
However,
the safety analysis
assumed
an initial boron concentration of 4000
ppm
and the plant procedures
did not limit the source of makeup water to a
source of such boron concentration.
The licensee's
investigation into these
issues
was discussed
in its Condition
Report/Disposition
Request
9-4-0171,
dated
March 17,
1994.
Condition
Report/Disposition
Request
9-4-0171 also discussed
two other reactor
physics
issues of concern that involved the following:
Technical Specification
3. 1. 1.3. restricted
moderator temperature
coefficient to be within the area of acceptable
operation
as specified
in the core operating limits report.
However. the safety analysis
assumed that the, moderator temperature coefficient for single
uncontrolled control element
assembly
bank withdrawal within the
deadband with control element
assembly calculators
was more
conservative.
By letter dated
June
3
~
1994, the licensee
submitted Licensee
Event
Report
(LER) 528;
529: 530/94-002-00.
The
LER reported
an April 22,
1994,
identification that three Technical Specification limiting conditions for
operation that would not ensure plant operation within the safety analysis
assumptions
as required
by 10 CFR 50.36.
The licensee's
letter indicated that
it had received
an extension to the original
LER report due date to
June 3,
1994.
The limiting conditions f'r operation in question
involved the
following:
~
Technical Specification 3. 1. 1. 1. requi red
a
1 percent
Modes 3, 4,
and
5 with all full-length control element
assemblies
fully
inserted.
However. the saf'ety analysis
and plant operating
procedures
required that the boron concentration
be maintained greater than the
boron concentration for hot full power with all rods out and equi librium
xenon with the reactor trip breakers
closed.
1,
i
h
-6-
Technical Specification 3.2.3. restricted
core azimuthal tilt to less
than
10 percent with core operating limit supervisory
system out of
service.
Howevers
the safety analysis
assumed that the tilt was less
than, or equal to,
3 percent with the core operating limit supervisory
system out of service.
The licensee's
investigation into core azimuthal tilt was also discussed
in
Condition Report/Disposition
Request
9-2-0326.
which was issued
on May 29,
1992.
On August 8.
1994. the licensee's
submitted
Supplement
01 to the
LER.
The
supplement
reported
a June
7,
1994, identification that
a fourth Technical
Specification limiting condition for operation that was also inadequate.
This
limiting condition of operation involved the following:
Technical Specification Table 4.3-1 required adjustments
to linear power
level, core protection calculator delta
power,
and core protection
calculator nuclear
power signals if they differed from the calorimetric
by an absolute difference of greater than
2 percent
when greater than
15 percent rated thermal
power.
However, the fuel vendor supplied core
protection calculator addressable
constants
that required the
calibration tolerance to be administratively restricted
below 30 percent
power.
The licensee's
investigation into these
issues
was discussed
in its Condition
Report/Disposition
Request
9-4-0338,
dated
May 20 '994.
On October
28 '994, the licensee
submitted
Supplement
02 to the
LER.
The
supplement
reported
a September
6,
1994, identification of the following:
The plant operating
procedures
for exercising control element
assemblies
in Modes 3, 4.
and
5 could allow conditions that violate assumptions
used in the subcritical control element
assembly
bank withdrawal
analysis.
The supplement
also discussed
a June
1991, discovery
by the fuel vendor of:
Two non-conservative
errors in the subcritical
neutron source strength.
Additionally. the supplement
stated that the licensee
had determined that the
previously reported
issues
involving Mode 6 boron dilution and the core
protection calculator power calculations
were not reportable.
The licensee's
investigation into these
issues
was discussed
in its Condition
Report/Disposition
Request
9-4-0641,
dated
September
6,
1994.
Following a review of the
LER and its supplements,
NRC decided to initiate a
t
followup inspection.
The primary inspection objectives for this inspection
were:
I
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To understand
the specific problems discussed
in the
LER as they related
to the safe operation of the plants
and, in particular, the licensee's
ractice of abandoning certain non-conservative
Technical Specification
imits and establishing
more conservative administrative limits that
were not been submitted for NRC staff review;
and
To gain
a general
overview of how well the licensee's
reload safety
analysis
process
was working.
The
NRC inspection of these
licensee
and fuel vendor identified issues
and the
licensee's
experience with reload safety analyses
is given below.
1.2
LER Issues
The issues
described
in the
LER and its supplements
were complex, often
inter-related,
and difficult to understand,
given thei r long-term histories.
The inspection of these
long-term outstanding
issues
revealed historical
management difficulties in ensuring priority consideration to resolving with
NRC technical
issues potentially impacting safe plant operation.
Although
generally untimely, engineering
decisions
related to revised core physics
limits were technically appropriate.
There were no technical
inadequacies
found in the licensee's
revised administrative limits that had supplanted
Technical Specification limits.
The
NRC inspectors'eview
of the
LER issues is separated
below into five
categories:
shutdown control element
assembly
bank withdrawal
and source
strength,
moderator temperature coefficient, core azimuthal tilt, core
protection calculator calibration,
and
Mode 6 boron concentration.
1.2. 1
Shutdown Control Element Assembly
Bank Withdrawal
and Source Strength
Inadvertent control element
assembly
bank withdrawal from a subcritical
condition is an anticipated operational
occurrence that can occur
as
a result
of operator error or hardware failure.
The event is analyzed in Chapter 15.4
of the updated final safety analysis report.
During subcritical conditions,
neutronic
feedback
mechanisms
are not significant because
the power generation
in the core in not large enough to cause
appreciable
changes
in fuel and
moderator
temperatures.
Consequently,
the event must
be terminated
by
a core
protection calculator trip or
a high logarithmic power level trip.
Acceptance
criteria for this event includes limitations on fuel temperature
and cladding
strain (controlled by fuel centerline temperature
limitation) and heat flux
(controlled by minimum departure
from nucleate
boi ling ratio).
In the original final safety analysis
report. the subcritical control element
assembly
bank withdrawal analysis with four reactor coolant
pumps running
assumed
the initial conditions to be those permitted
by the Technical
Specification limiting conditions of operation.
Automatic protection
was to
be provided by the high logarithmic power level trip at 0.8 percent
rated
thermal
power.
Core protection calculators
were designed to provide
a reactor
trip when required for anticipated operational
occurrences
and postulated
I
-8-
accidents
when initiated from a power level greater
than the core protection
calculator operating
bypass setpoint.
For conditions with less than four
pumps in operation,
the analysis
took credit f'r the core
protection calculator trip bypass
removal
upon attaining
a
1 percent
rated
thermal
power as providing
a trip.
Both analyses
concluded that the results
were acceptable.
the specified acceptable
fuel design limits were not
exceeded.
and General
Design Criteria 20 and
25 (protection system functions
and protection system requirements
for reactivity control malfunctions,
respectively)
were satisfied.
On October 9.
1987.
NRC approved the licensee's
January
23,
1987,
request for
Technical Specification
changes
involving shutdown margin and. in particular,
the high logarithmic power level trip and core protection calculator
bypass
setpoints
~ which were changed
from 0.8 and 1.0 percent of rated thermal
power,
respectively,
to 0.01
and 0.0001 percent of rated thermal
power, respectively.
The revised analysis for the subcritical control element
assembly withdrawal
with four reactor coolant
pumps
was incorporated into the updated final safety
analysis report,
but the subcritical control element
assembly
bank withdrawal
with less
than four reactor coolant
pump operation
was not.
The updated final
safety analysis
report did not address
core protection calculator operability.
Core protection calculators
are required to be operable in Modes
1 and 2,
only.
The lack of Technical Specification operability criteria for the core
protection calculators
in Modes 3, 4,
and
5 was not recognized.
As given in
Condition Report/Disposition
Request
9-4-0171,
Item 2, credit for core
protection calculator trips was beyond the design basis of the core protection
calculators.
The licensee's
fuel vendor
informed the licensee of an error in the analysis
of subcritical neutron source strength
on May 24.
1991.
The fuel vendor
had
determined that the actual subcritical
neutron source strength
was smaller
than previously predicted.
(For the subcritical control element
assembly
bank
withdrawal event,
a smaller neutron source strength will increase the amount
of neutron multiplication prior to reaching the trip setpoint, therefore,
resulting in a higher
power spike than would have occurred with a higher
neutron source strength.)
The fuel vendor indicated that the error
was not
a
safety concern
because
the event analysis
contained
enough conservatism
in the
assumed initial conditions to demonstrate
results within the licensing basis
when the corrected initial subcritical
power level was used.
In addition, the
fuel vendor determined that the error was not reportable
by the requirements
of 10 CFR Part 21.
The letter stated that while a reanalysis
was being
performed'aintaining
boron at or above hot full power, all rods out,
equi librium xenon concentration
would resolve the issue.
The licensee
issued
a night order on May 29,
1991, to ensure this measure.
I ~
f
I
-9-
After the fuel vendor letter of May 24.
1991, it was determined
by the
licensee that the error was reportable
under
but further
investigation
was necessary.
Interim corrective action was being developed to
ensure that plant operation
was maintained in an analyzed configuration.
Actions to be taken were listed in Condition Report/Disposition
Request
3-1-0007,
issued
June
6
~
1991.
On June
17.
1991, the licensee's
fuel vendor informed the licensee of the
final results of the subcritical
source strength analysis.
The fuel vendor
had determined there were actually two errors in the generic calculation
used
to calculate the source term.
The erroneous
subcritical neutron source term
was non-conservatively
high by
a factor of about
5000.
The fuel vendor wrote
that. the licensee's
boron dilution restriction or the core protection
'calculator trip would provide adequate
protection to account for the errors.
On June
24.
1991, it was determined
by the licensee that the source term error
was not reportable despite the earlier statement
in Condition
Report/Disposition
Request
3-1-007.
The reasoning
was that it was not
a
condition that would significantly compromise plant safety.
The licensee
received
a reanalysis of the subcritical control element
assembly
bank withdrawal event dated October
8,
1991;
however,
inasmuch
as the licensee
was relying on its fuel vendor for complete reload safety analyses
at that
time no licensee detailed
review of this document took place.
Yet, there
clearly was
a plant condition for which the fuel vendor gave credit to
operating restrictions that the licensee
was free to change without
consultation with the fuel vendor.
In 1992,
a groundrule
change
was initiated that split the subcritical control
element
assembly
bank withdrawal analysis into a shutdown
bank and
a
regulating
bank analysis.
The shutdown
bank analysis
used the boron
concentration greater than hot full power, all rods out,
equi librium xenon
concentration
and the regulating
bank analysis
used the high logarithmic power
level trip as automatic protection.
Core protection calculators
continued to
be used for generating
the trip when less than four reactor coolant
pumps were
running.
On May 18.
1994.
changes
to the Technical Specification were proposed to make
the limiting condition for operation for shutdown margin more restrictive by
requi ring the boron concentration to be greater than hot full power, all rods
out.
equi librium xenon concentration.
On September
6,
1994 'ondition Report/Disposition
Request
9-4-0641
was
written identifying that control element
assembly
exercising in Modes 3, 4,
and
5 could violate the assumptions
in the subcritical control element
assembly
bank withdrawal analysis.
The inspectors
noted that
a special plant
review board met in a timely manner (the
same
day as the Condition
Report/Disposition
Request
9-4-0641
was initiated) to discuss
the latest
information on the control element
assembly
issue.
Since the licensee's staff
1
~
i
-10-
had not finalized
a proposed solution, the board took positive action to
require any future control element
assembly withdrawal to be reviewed
by the
board prior to its occurrence.
In LER 528;
529; 530/94-020-404 it is stated
that Technical Specification
changes
on this subject were expected to be
submitted to the
NRC by December
30,
1994.
On December
16,
1994,
licensee
representatives
notified the inspectors that they had informed the Office of
Nuclear Reactor Regulation Project
Manager that this date
was slipped to
January
30 '995.
There were missed opportunities to identify and correct deficiencies.
The
issue of subcritical control element
assembly
bank withdrawal identified in
the March
17 '994.
Condition Report/Disposition
Request
9-4-0171
was
originally recognized
by the licensee's
reactor engineering staff in November
1993.
In November
1993. during the review of a
new operating procedure.
a
discrepancy
was identified that the fuel vendor's
guidance to maintain boron
restrictions or core protection calculator operable
was insufficient.
The
licensee's staff questioned that both the boron restriction
and the core
protection calculators
were necessary.
The issue
was telephonically referred
to nuclear fuel management
personnel,
without initiating a condition
report/disposition
request.
Procedure
Revision 2. effective
November 2.
1992.
was the effective procedure describing the condition
reporting process.
The procedure
described
a process
for the identification,
documentation'nd
evaluation of conditions that
may adversely effect the safe
operation of the plants.
Section
1. 1.3 of the procedure identified that
conditions which may adversely effect the safe operation of the plant were
human errors;
procedure deficiencies
and technical
inadequacies;
use or
generation of incorrect
and inadequate
documents
such
as specifications,
procedures'r
instructions;
and conditions that could result in reports to
external
agencies.
Section 3. 1.2 of the procedure
required that if the
condition described
required
immediate action
or
may have adverse
or immediate
impact on the operation of plant systems
or equipment.
then the originator
shall initiate any requi red immediate actions
and complete the condition
report/disposition
request
in accordance
with the condition report/disposition
request instructions to complete the applicable sections of the form as soon
as practical.
By an internal
memorandum
dated February
21,
1994, nuclear fuel
management
personnel
formally specified that both the boron restriction
and
core protection calculators
were necessary.
A condition report/disposition
request
was not initiated at that time, contrary to the effective version of
Procedure
~ Revision 3, issued
on January
3
~
1994.
A condition
report/disposition
request
was finally initiated on March 17.
1994.
Subsequently.
the mode change checklists in Procedure
"Nuclear
Administrative and. Technical
Manual." were changed to put in place the new
requi rements
on March 19,
1994.
A request for a Technical Specification
change
was not initiated at this time, but according to Condition
Report/Disposition
Request
9-4-0171.
a change will be proposed.
This issue of
not initiating a condition report/disposition
request
in a timely manner
and
in accordance
with the appropriate categorization
is
a violation (528:
529;
530/9435-01).
l
e
On November 29.
1994. the licensee's
representatives
presented
to the
inspectors
a partially drafted condition report/disposition
request
which
requested
raising the sensitivity to the need to initiate a condition
report/disposition
request expeditiously,
investigate the time allowable
between discovery
and initiation of a condition report/disposition
requests
determine whether training is required.
and determine whether the issue is
applicable to organizations
outside nuclear
fuel management.
1.2.2
Moderator Temperature Coefficient
The fuel vendor notified the licensee that analysis for a single uncontrolled
control element
assembly withdrawal within the control element
assembly
motion
inhibit/prohibit deadband of the control element
assembly calculators,
with
the calculators
assumed
a moderator temperature coefficient value
that was more conservative
than the Technical Specification value.
The reload
groundrules
assumed
the most positive value of moderator temperature
coefficient to be 0.0 pcm/'F at 60 percent of rated thermal
power, while the
Technical Specification limit, as reflected in the core operating limits
report.
allowed
a more positive value of + 2.0 pcm/'F.
The Technical
Specification basis stated that the limitation on moderator temperature
coefficient was provided to ensure that the assumptions
remained valid through
the each fuel cycle.
The inspectors
confirmed that the licensee
had implemented
changes
to the Palo
Verde Nuclear Generating Station core operating limit report for moderator
temperature coefficients.
The current Technical Specifications for all units
refer operators
to the core operating limit reports that
now provide more
restrictive limits that agree with the event analysis
assumptions.
The
inspectors
noted that the licensee's
corrective action system indicated these
changes.
The inspectors
reviewed the surveillance
requirements
of the Technical
Specifications
and verified that surveillance testing
was requi red to be
performed three times during each fuel cycle to ensure that assumptions
used
in accident
and transient analysis
remained valid through the cycle.
According to statements
in the corrective
action document,
Condition
Report/Disposition
Request
9-4-0171. startup physics testing
was used to
verify core operation within the assumption of the safety analysis.
The
inspectors
reviewed
one startup physics testing record provided by the
licensee
(Unit 2 Cycle 5) and verified the data
was acceptable.
The inspectors
believed that this issue of moderator temperature coefficient
discrepancy
could have
been identified and resolved
much earlier than it was.
There were several
opportunities for identification during the development of
previous core reload analyses
and
10 CFR 50.59 reviews of the analyses
by the
fuel vendor
and the licensee that should have identified this discrepancy.
I
l
J
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1.2.3
Core Azimuthal Tilt
Prior to late 1992. the core accident
analyses
design basis limit for core
azimuthal tilt. which was consistent with the groundrules,
was inconsistent
with the Technical Specification limit. Specifically, the analyses
assumed
that with the core operating limit supervisory
system out-of-service there
would be
a 3 percent tilt (rather than the
10 percent
allowed by Technical
Specification).
In response
to this discrepant
findings the licensee
issued
Condition Report/Disposition
Request
9-2-0326
on Hay 29,
1992.
The March 17,
1994. Condition Report/Disposition
Request
9-4-0171 relied on
the
1992 reportability determination that this issue
was not reportable.
In
reaching that
1992 determination.
the licensee's
logic had been
(1) the
problem would only occur for
a limited time in a cycle or (2) administrative
procedural
controls were in place for the problem.
The licensee's
staff
committed
a large amount of time in reaching its determination that the issue
was not reportable.
According to Condition Report/Disposition
Request
9-2-0326, its staff reviewed hundreds of records in reaching its decision.
The licensee's
fuel vendor was requested
to provide input to the licensee's
decision
making on how to resolve the issue.
A July 1.
1992,
response letter
from the fuel vendor
stated:
The downside risk to the Technical Specification
change
approach
would be
NRC review and questions
on the cause for the current
measured tilt and generic implications for other
C-E plants with
digital protection systems.
This guidance
from the fuel vendor was unquestionably
inappropriate.
It
minimized the significance of a potentially generic issue
and contributed to
the licensee's
uncertainty in how to enact
a satisfactory resolution to its
azimuthal tilt quandry.
Ultimately. the licensee
decided to resolve this issue
by submitting
a
Technical Specification
amendment
request
on January
4,
1994.
The request,
approved
by
NRC on November
3
~
1994,
lowered the Technical Specification limit
to 3 percent.
Additionally, the
licensee
changed the appropriate
procedures
to require that for instances
when the core operating limit supervisory
system
is in service but both of the control element
assembly calculators
are
inoperable that require
a power reduction to less than
50 percent of rated
thermal
power if core azimuthal tilt exceeds
3 percent.'lthough
the
licensee's
decision
was appropriate,
this issue,
a potential generic issue,
should
have
been corrected in a more timely manner.
The inspectors
discussed
with the licensee's
representatives
plant operation
prior to late November
1992 when core azimuthal tilt had been limited to
10 percent.
The licensee's staff identified historical occurrences
where
plant tilt had exceeded
the 3 percent
design basis.
The inspectors
questioned
the logic of the licensee's
reportabi lity determination.
inasmuch
as the issue
of operation outside the design basis
needed to be quantified for the
significance of those occurrences.
For these
occurrences.
the licensee's
t
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0
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staff explained that the occurrences
were of short duration,
thereby the
probability of asymmetric conditions that could have perturbed
core azimuthal
tilt was small.
The summary information reviewed
by the inspectors
showed
that for those occurrences,
the Technical Specification action statements
had
not been violated (which if they had would have necessitated
a separate
reporting to NRC).
After returning to the Regional office, the inspectors
continued the
inspection of this issue of outside-design-bases
occurrences
with the
Office for Analysis
and Evaluation of Operational
Data to determine whether
the occurrences
should
be considered
reportable
pursuant to
10 CFR 50.72(b)(1)(ii)(B) or 50.73(a)(2)(ii)(B).
Both specified regulations
refer to reporting requi rements for a condition that is/was outside the design
basis of the plant.
It was subsequently
determined
from discussions
and
review of related written guidance (including examples) that the licensee's
occurrences
were not reportable
under the specified reporting requirements.
It was concluded that these
two specific reporting requi rements
were
associated
with higher-tier design basis associated
with fuel protection.
For
the Palo Verde Nuclear Generating Station,
the minimum departure
from nucleate
boil'ing safety limit of 1.24 or fuel rod heat generation
rate safety limit of
21 kW/ft that had not been violated were subject to these reporting
requirements.
This finding was provided to licensee
representatives
on
December
16.
1994, in a telephone
conference.
During the review of this issue,
the inspectors
were informed by the
licensee's staff that they did not consider matters
such
as core azimuthal
tilt accident analyses,
which had led to plant operational limits such
as
3 percent core azimuthal tilt, to constitute design
bases.
The inspectors
expressed
disagreement
with that viewpoint, which contradicted their
understanding
of reactor
safety analyses
requirements,
as set forth in staff
documents
such
as the standard
review plan.
1.2.4
Core Protection Calculator Calibration
In 1988 'he licensee's
fuel vendor identified the need for more restrictive
requirements
for adjustments
to linear power level, core protection calculator
delta
T power,
and core protection calculator nuclear
power when differing
from calorimetric calculation
by more than +/- 2 percent.
The licensee
put
administrative controls in place at that time, but did not request
a Technical
Specification
change.
The administrative controls included changes
in the
core protection calculator
setpoint analysis
and power calibration procedures
to eliminate the potential for non-conservatism
in core protection calculator
calculations of departure
from nucleate
boi ling and linear heat rate.
The inspectors
asked to review the
10 CFR 50.59 evaluations
performed for
these administrative controls
and the associated
reportabi lity determination
for this issue.
but these
documents
were not available.
I
S
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In Harch 1989. the licensee's
fuel vendor, 'in reviewing the licensee's
administrative controls.
recommended that the appropriate
Technical
Specification revision should
be submitted to "eliminate the need for this
complex interim approach."
In making
a decision whether to request
an
amendment.
the licensee's staff evaluated other similar vintage Combustion
Engineering nuclear
steam supply system licensee
correspondence
to change
related Technical Specification limits.
For
a variety of reasons,
several
of
those submittals
were unacceptable
to the
NRC staff,
and the licensee's staff
inappropriately derived
an
NRC staff viewpoint from those transactions
rather
than pursuing its own Technical Specification
change.
The use of administrative controls that are more restrictive than Technical
Specification limits is
a
common practice to protect against violating the
Technical Specification limits.
However, the abandonment of Technical
Specification limits for the use of more restrictive administrative limits to
control safety analysis
assumptions
as
a prolonged or routine practice instead
of requesting
NRC review and revision of Technical Specification limits in
accordance
with 10 CFR 50.36 was not prudent.
As stated in Section
1. 1 of this report, the licensee
had recently determined
on June
7,
1994, that this issue
was reportable.
That determination of
reportability had been initiated as early as
May 24,
1994, during
a revision
to Condition Report/Disposition
Request
9-4-0338.
The reportabi lity
determination
was again confi rmed in an internal
memorandum
dated
June 9.
1994 'herein it was stated that the report was to be made in
Supplement
1 to LER 528:
529; 530/94-002-00.
Supplement
1 to LER 528;
529;
530/94-002-00.
though.
was not issued until August 8,
1994.
Consequently,
the
licensee's
report was technically more than
a month overdue.
However,
Supplement
2 to LER 528:
529: 530/94-002-00,
issued
on October 28,
1994,
stated that this issue involving the core protection calculator
power
calculations
had
now been determined
not to be reportable.
During discussions
with the inspectors,
the licensee's
representatives
explained that their best recollection
was that they had received telephonic
approval
from the Walnut Creek Field Office for an extension to the reporting
timeliness for Supplement
1 to LER 528;
529: 530/94-002-00;
however,
the
licensee
and
NRC personnel
could not identify a record of the extension
approval.
This reportabi lity review was also confusing to the
NRC inspectors
because
on November 15,
1994, the licensee's staff made
a presentation
to the
inspectors,
during which a handout indicated that this issue
was reportable
and did not acknowledge the prior October 28,
1994, determination.
The inspectors
believed that the licensee's
implementation of its
reportabi lity process's
indicated
by this example
and its frequent
requests
to the
NRC for time extensions
related to licensee
correspondence,
was in need
of significant improvement.
I
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0
1.2.5
Mode 6 Boron Concentration
-15-
The initial reactivity analysis
done by the licensee's
fuel vendor
assumed
that the boron concentration for the
Mode 6 boron dilution analysis
was
4000
ppm.
based
on the Technical Speci.fication requirement for the refueling
water tank stored water.
This assumption
was less conservative
than the
2150
ppm or K-effective less than 0.95 requi red by Technical Specifications
for refueling.
This error was not discovered until the Unit 3, Cycle 5 analysis
was being
performed.
After discovery of the error,
a reanalysis of Unit 3, Cycle 5 was
performed
and it verified that the source
range monitoring setpoint ratio
of 2.2 was valid, but the core operating limits report requi red revision to
Table 5.
A reanalysis
of Units
1 and 2, Cycle 5 analyses
revealed similar
results.
The licensee's
review of plant operations identified two situations in which
the assumptions
of greater
than or equal to 4000
ppm was violated for an
extended
time.
The first situation
was in 1988,
where for approximately
17 days boron was allowed to decrease
to 2275
ppm.
The second
case
was
between April
1 and
May 13,
1988,
when the boron concentration
was
as
low as
3700
ppm.
These situations.
however,
met the 2150
ppm limit in Technical Specification 3.9. 1, although they did not meet the safety analysis
assumption
of 4000
ppm.
Apparently, the licensee's
fuel vendor never incorporated the assumption into
the reload design groundrules for the licensee's
review.
The licensee
deferred to its fuel vendor's
experience,
and apparently did n'ot question the
assumptions.
This discrepancy
existed for a long time.
It had the potential
to be discovered at each of 14 reloads
before it was found.
1.2.6
Conclusion
Closed
LER 528
529
530/94-002
and its Su
lements
01 and 02:
Technical
S ecification Limitin Conditions for o eration that would not ensure
lant
o eration within the safet
anal sis
assum tions
as re uired
b
According to the discussion
above, this
LER is closed.
1.3
Use of Site-Wide Administrative Processes
During the licensee
evaluation of issues
leading
up to the issuance
of
LER 528;
529: 530/94-002-00,
licensee
personnel
utilized site-wide
administrative processes
to evaluate
and disposition discrepancies.
The
inspectors
followed up and assessed
the use of the site-wide administrative
processes
for corrective action tracking and procedure revision.
l
-16-
1.3. 1
Corrective Action Tracking Process
The inspectors
determined that
LER 528;
529: 530/94-002-00
resulted
from
licensee
and fuel vendor identified findings that were reported in three
different corrective action documents.
Condition Report/Disposition
Requests
9-4-0171,
9-4-0338,
and 9-4-0641 were initiated in response
to the
findings.
Additional condition report/disposition
requests
were reviewed,
but
the inspection
focussed
on the effectiveness
of these three primary condition
report/disposition
requests.
The inspectors
reviewed the licensee's
guidance for the corrective action
system,
contained in Procedure
"Condition Reporting," Revision 4.
The procedure
contained all the administrative information necessary
to
identify a problem,
perform assessment
or evaluate,
classify (for
significance),
determine corrective action,
and track corrective action.
The
inspectors
referenced
the three previous revisions of this procedure to
determine if the licensee's staff had handled the corrective action process
properly in light of many recent significant process
changes.
Condition Report/Disposition
Request
9-4-0171
was issued
when the licensee
determined that
some Technical Specification limiting conditions for operation
did not ensure that plant operation
was conducted within analyzed conditions.
The specific limiting conditions for operation were related to shutdown
margin, core protection calculator operability, moderator temperature
coefficient, refueling boron concentration,
and core azimuthal tilt.
The inspectors
experienced difficulty in following corrective actions
because
of specific unique practices
associated
with tracking.
For example,
the most
current version of Condition Report/Disposition
Request
9-4-0171 contained
a
list of five corrective actions that had already
been completed.
Howevers
these five items did not appear in the database.
so the inspectors
asked for
the closure references
stated
in the Condition Report/Disposition
Request
9-4-0171.
Likewise, Corrective Action Items
9 and
10 to be completed
were not in the database.
The inspectors
noted this to a licensee
representative
who stated that corrective actions
completed before
a condition
report/disposition
request
was placed into the database
were not entered into
the database.
Additionally, corrective actions initiated after
a condition
report/disposition
request
was placed in the database
were often issued
as
separate
action items.
Searches
for documents to substantiate
closure of some
corrective actions that were not in the database
proved to be long and
exhaustive.
The inspectors
opted not to challenge the licensee's
resources
and did not insist on seeing all documentation.
When Condition Report/Disposition
Request
9-4-0171
was written. Revision
2 of
Procedure
90AC-OI04 was in effect.
The inspectors
noted that Condition
Report/Disposition
Request
9-4-071
had been classified
as Category
4 with a
root cause analysis
requested.
The inspectors
reviewed the classification
guidance of Revision
2 and noted that the guidance contained
a set of
questions
to determine if a condition report/disposition
request
was to be
Category
3.
Question
Number
4 read,
"Does the condition involve
k
~
I
I
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administrative,
procedural
~ or operational
errors that demonstrate
a
fundamental
misunderstanding of'r noncompliance with operational,
regulatory,
or nuclear safety requirements.
Lsic ?]"
Based
on this guidance.
the
inspectors
determined that the condition report/disposition
request
had been
misclassified.
However. after further discussion with quality assurance
management
and review of the procedure,
the inspectors
were not sure that
an
apparent misclassification would significantly affect final corrective action.
Additionally, further review of the current procedure revision revealed that
a
condition report/disposition
request classification of Category
1 (potentially
significant) required
some type of followup assessment
or evaluation,
such
as
root cause determination
or team investigation.
Unlike previous revisions,
management
had more flexibilityin determining the type of followup evaluation
to be performed.
Condition report/disposition
requests
classified
as
Category
2 (nonsignificant) did not requi re the performance of formal followup
assessment
or evaluation.
The inspectors
believed that the safety issues
related to Condition Report/Disposition
Request
9-4-0171
had been adequately
addressed.
Condition Report/Disposition
Request
9-4-0338 was issued
when the licensee
identified that adjustments
to nuclear
power instruments
and signals
were
required
when the absolute difference from calorimetric calculation
was
greater
than
2 percent,
but analysis
assumed
a lesser
difference
(+2 percent,
-0.5 percent)
below 30 percent
power.
The inspectors
reviewed the current disposition of Condition Report/
Disposition Request
9-4-0338 'hich remained
an open corrective action
document.
The inspectors
experienced difficulty in determining
how the
licensee
applied immediate.
short-term corrective action.
Eventually, it was
determined that the short-term corrective
action was to continue
administratively controlling instrument calibration to the stricter allowed
difference.
This decision
was
made prior to entering the condition
report/disposition
request into the database
and the action needed to do this
was not evident in the database.
To accomplish this, it was necessary
to
maintain requirements
found in two procedures.
Subsequently,
according to
a
licensee
representative.
the licensee's
procedure
revision process
used in
conjunction with the corrective action tracking system would prevent
inadvertent elimination of the corrective action by freezing certain
requirements
in:
Procedure
"Adjustable Power Signal Calibration," Revision 3;
and
~
Procedure
"Power Calibration," Revision 3.
Because of time constraints,
the inspectors
did not verify the licensee
representative's
assertion
regarding the preservation of corrective action.
Additionally. according to tracking system information. all required
corrective actions
had been completed
and closure
was pending final review.
'
0
-18-
These tracked actions
included
a 10-day draft evaluation.
a complete
evaluation,
a reportability review,
and
a nuclear
assurance
review.
The
inspectors
believed that the licensee
was adequately
addressing
safety issues
developed
by Condition Report/Disposition
Request
9-4-0338.
Condition Report/Disposition
Request
9-4-0641
was initiated to address
a
condition where exercising control element
assemblies
while shutdown would
exceed the assumed
conditions for analysis of subcritical control element
assembly
bank withdrawal.
The inspectors
reviewed the current disposition
and
corrective action tracking of Condition Report/Disposition
Request
9-4-0641.
The following corrective action was proposed:
Draft Evaluation.
Root Cause
Evaluations
Nuclear Assurance
Review for Corrective Action Adequacy,
Reportabi lity Review,
Regulatory Affairs Complete
LER.
and
Nuclear Assurance Verification of Corrective Action'ompletion.
According to the tracking system,
only the root cause evaluation
and
verification of corrective action completion remained
open.
The inspectors
found it a poor practice that the various corrective actions could be
determined
and apparently finalized, with the root cause evaluation
determination
not final.
To assure that the safety concern
had been addressed,
the inspectors verified
the completion of the specific corrective action to address this concern.
This corrective action was implemented in Procedure
"CEA
Exercising in Modes 3, 4,
and 5." Revision 1.
The procedure
contained
precautions
and limitations to assure
adequate
shutdown margin would exist by
requi ring one control element
assembly calculator
and three core protection
calculators to be operable prior to exercising the control element
assemblies.
In additions
was required to contain
a boron
concentration to provide shutdown margin for conditions of hot full power, all
rods out,
and equi librium xenon for current core burnup.
The inspectors
agreed with the licensee
conclusion that these
cor rective actions
adequately
addressed
the safety issue.
The inspectors
followed up on the licensee's
handling of the issue associated
with the fuel vendor's discovery of errors in the calculation of subcritical
neutron source strength.
The inspectors
inquired if, at the
May 24,
1991,
report time. there
was
a site-wide process
for handling vendor technical
information arriving at the site.
Licensee
personnel
stated that in May 1991,
a program was in place to assure
the correct handling of this type of
information.
and the program was ongoing.
However, fuel vendor information
had always
been
exempted
from the requirements of the program.
Therefore,
this letter had been sent directly to the Manager,
Nuclear Fuel
Management,
and never entered into the site-wide administrative information processing
system.
I
I
I
I
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The inspectors
asked representatives
of nuclear fuel management
how this
correspondence
had been treated within that organization.
According to
documentation
provided.
Problem Resolution
Sheet
0001478
was initiated on
May 29 '991.
The problem resolution sheet
was the licensee's
primary
corrective action document prior to the origination of the condition
report/disposition
request.
The problem resolution sheet indicated that the
shift technical
advisor
and the assistant shift supervisor
required
a
reportabi lity determination.
The plant manager
assigned
an investigation
director on May 30.
1991.
and the issue
was assigned
an investigation
Category 4.
There was
no further apparent
actions related to the problem
resolution sheet
issues.
Howevers
the inspectors
were provided
a copy of
Condition Report/Disposition
Request
3-1-0007 which had been initiated on June
6,
1991.
According to the condition report/disposition
requests
the concerns
originally stated in the problem resolution sheet
were transferred to and
addressed
in the condition report/disposition
request.
After the events
described
above,
the licensee
received the second letter
related to subcritical
source neutron strength
on June
17,
1991.
The letter
further reported the results of the fuel vendor 's review of the current
analysis
and the licensee's
exitant procedures
indicated sufficient
conservatism with current operating practices.
The inspectors
asked licensee
representatives
how conditions
and procedures
were identified and controlled
to provide the needed
conservatism.
Licensee
personnel
referred the
inspectors to Condition Report/Disposition
Request
3-1-0007,
which indicated
that one of the actions to fix the problem was to allow the plant to continue
to operate
by taking credit for current operating procedures.
This action was
stated to be complete.
According to the documentation
reviewed,
Condition
Report/Disposition
Request
3-1-0007
had been closed.
The inspectors
again
asked what procedures
were identified and
how were they controlled.
Just
prior to the end of the inspection,
the inspectors
were provided with a
tracking system report
and licensee Letter 261-00243-MAH regarding the closure
of Problem Resolution
Sheet
1478.
The letter stated that
13 procedures
had
been
changed to address
the problem of the error in the calculation of the
subcritical
source term.
The tracking system report indicated that the
procedures
referenced
in the letter were indeed disposition documents for the
problem resolution sheet.
There was insufficient time during the remainder of
the inspection to assess
the role in each procedure
toward satisfying the
corrective actions.
The inspectors
considered
the licensee's
corrective action tracking process
to
be weak in that it allowed unnecessary
vulnerability to missing necessary
correction actions.
1.3.2
Procedure
Change
Process
The inspectors
noted that procedure
steps to meet commitments
were not
identified within the text of the procedures.
The inspectors
discussed
with
licensee
personnel
the subject of preservation of commitments that were
'I
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implemented through procedure revision.
During these discussions.
inspectors
gained
an impression that
some Nuclear Fuel
Hanagement
personnel,
including
supervision.
were not aware of the need to have this element in the procedure
revision process,
or
how the element
was satisfied.
The inspectors
reviewed licensee
Procedure
"Review and Approval of
Nuclear Administrative and Technical
Procedures,"
Revision 5, which provided
the guidance for revising safety-related
procedures.
Procedure
Step 3.2.2
requi red the performance of a corrective action tracking system database
search.
Appendix A to this procedure
provided the procedure action cover
sheet which contained
a statement that
a corrective action tracking system
database
search
was required for all procedure actions.
Procedure
Appendix
H
provided the instructions for
a database
search
and report printing.
The
inspectors
concluded there were programmatic
requirements
in place to assure
that commitments
implemented within procedures
would not be eliminated
by
procedure revision.
However, there did not appear to be any safeguards
to
assure that the database
search
was actually performed
and correctly
interpreted.
The inspectors
believed there to be
a vulnerability to eliminate
commitments
implemented in licensee
procedures
inadvertently.
1.4
Reload Safet
Anal sis
Ex erience
1.4. 1
Reload Safety Analysis History
The licensee
received
NRC approval
on June
14,
1994, to conduct portions of
its own reload safety analyses
using Combustion Engineering
computer codes.
The specific computer
codes that were approved for licensee
use included codes
for the analysis of physics.
thermal-hydraulics,
fuel performance,
and certain
and accidents.
Subsequent
to this approval,
the licensee
has
maintained
a partnership
agreement with its fuel vendor
as
a joint continuing
effort to provide reload safety analyses.
Previously in 1992, the licensee
took control over the groundrules
development.
The groundrules
document
provides
a reload design interface with the fuel vendor
and establishes
design
assumptions
used in reload analyses.
The licensee
was using
a team composed of fuel vendor
and licensee
personnel
(2/4, respectively) to perform reload safety analyses,
except for Unit 1,
Cycle 5 for which the licensee did not have resources
to support the reload
safety analyses
and utilized all fuel vendor
personnel
to develop the reload
safety analysis report.
From the brief overview. the licensee's
reload safety
analysis
reports
were found to be similar or somewhat greater in content
and
scope of detail to that provided by the licensee's
fuel vendor to other
licensees.
As part of the recent reengineering
process,
the licensee
recognized
a
weakness
in that they did not have
a high-level tier procedure that governed
the various involved procedures
to ensure that all plant changes
and concerns
about reload safety analyses
were captured for subsequent
considerations
for
)
I
( I
-21-
unit cycle bases
documents (i.e.. Technical Specifications,
updated final
safety analysis report,
and groundrules).
The license representative
stated
that they were working toward the development of this master
procedure
issuance
next year.
The licensee's
representative
stated that it intended to begin using the
NRC-approved
methodology for conducting reload safety analysis in the fall of
1995.
Until that time. they planned to continue to rely on assistance
from
its fuel vendor under the partnership
agreement.
It appeared
that reload safety analysis
process
was working well.
The
licensee's
intention of improving the controls over groundrules
was
a positive
idea.
The ongoing partnership
arrangement
with the fuel vendor
had increased
the licensee's staff's fuel engineering analytical capabilities
and had
brought both the fuel vendor
and the licensee's
attention
on specific
analytical
issues.
In short.
a better technical
coverage of issues
has
occur red.
Some examples of this better coverage
was evident in the discovery
of problems
as documented
in certain condition report/disposition
requests.
It was understood that it was the licensee's
goal to become
independent of
your fuel vendor for certain reload safety analysis'ut
the inspectors
had
the strong impression that the licensee did not have
an adequate
resource
base
in nuclear fuel management
to sustain three unit reload safety analysis
demands.
Whether or not that transition is made is the licensee's
prerogative,
but for the interim, the joint effort appears
to be an
improvement over the pre-1993 practices for conducting reload safety analysis.
1.4.2
Fuel
Thermal
Performance
Analysis
The inspectors
performed
a brief review of certain inputs to the Unit 2,
Cycle 5, fuel thermal
performance
analysis
(Analysis No. A-PV2-FE-0008,
dated
December
14.
1992).
The analysis
was performed using the FATES-3A version of
the C-E fuel evaluation
model.
The inputs were used to generate
information
necessary
to support the reload safety analysis that specified acceptable
fuel
design limits were satisfied,
such
as
peak fuel rod internal pressure
less
than normal reactor coolant system operating pressure
and fuel rod power less
than
21 kW/ft.
The analytical results
were generated
for the hot and average
rods for design parameter for Batches
B,
E,
F,
and
G and incorporated the
highest radial
peak over the entire Cycle 5 burnup range.
The analysis
established
that acceptable
results
were obtained.
The inspectors
considered
the analytical
analyses
to have been reasonable
and identified no problems.
1.4.3
10 CFR 50.59 Evaluations of Fuel
Design
Changes
The inspectors
reviewed the
10 CFR 50.59 safety evaluations for fuel design
changes
that were implemented during the most recent refueling outage for each
of the units.
The licensee's
representative
identified and provided three
such evaluations:
one for Unit 2
~ Cycle 5, operation
dated June
23,
1993:
one
for Unit 1. Cycle 5. operation dated
November
19 '993:
and one for Unit 3,
Cycle 5
~ operation
dated
March 9.
1994.
The design
changes that were
evaluated for each unit were not identical'ut included changes
such
as
a new
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"value added fuel pellet."
a new burnable poison absorber pellets
a
new low
volume fuel rod plenum
springer
the elimination of the fuel rod upper alumina
spacer
disks
an upgrade to the Inconel spacer grid assembly to improve debris
entrapment,
and the use of laser welding instead of tungsten inert gas welding
to produce smaller weld nuggets.
thereby,
reducing pressure
drop coefficients.
The inspectors
considered
the evaluations of the changes to be brief,~with an
almost complete reliance
on the fuel vendor's analytical results that were
obtained with computer
codes previously approved
by the
NRC and referenced
in
the reload safety analysis.
This was surprising
because
the licensee
was
making efforts to become
independent
in its capability to perform reload
safety analyses.
Nevertheless'he
inspectors
did not identify any specific
problem with the design
changes that could have
been considered
as
an
unreviewed safety question.
The inspectors
also reviewed the March 18,
1994,
and October
13.
1994.
10 CFR 50.59 reload safety analyses
for the Unit 2, Cycle 5
tube plugging mini-outages.
A necessary
change
from the assumed
number of
plugged
tubes
was that the peak linear heat generation
rate
needed to be restricted to 13.2 kW/ft, rather than 13.5 kW/ft previously used.
The licensee
implemented this reduction
by changing the core oper ating limit
supervisory
system addressable
constant
and the core operating limits report
value.
In addition, the licensee
requested
and received
a change to the
Technical Specifications to use the 1985 Evaluation
Model for the large break
loss-of-coolant
accident
for the generation of the core operating limits
report information.
The inspectors
inquired of the licensee's
representatives
as to whether the
licensee
had
made
any fuel assembly or control element
assembly
changes
for
any of the most recent unit refueling outages after the associated
reload
safety analyses
had been completed.
The licensee
representatives
stated there
was
one change
made to the Unit 1, Cycle 5, fuel loading scheme that occurred
as
a result of fuel vendor concerns
about fuel rod fretting in 8 specific fuel
assemblies
located
on the core periphery.
Subsequent
to the fuel vendor's
(August 4,
1993. Letter V-93-204), the licensee
decided to
replace the 8 fuel assemblies.
The licensee
response letter of August 4,
1993.
accepted
the fuel vendor recommendations.
Subsequently,
the fuel
vendor's letter of August 19,
1993. provided
a new fuel loading pattern.
The
licensee's
representatives
indicated that
a
10 CFR 50.59 review of this change
was not necessary
inasmuch
as the reload safety analysis
was not complete
and
approved at the time of notification of the
recommended
change.
The
inspectors
agreed with that viewpoint.
No discrepancies
were identified during the reviews.
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1.4.4
Reload Analysis Training
The
NRC approved
a topical report describing the transfer of reload analysis
technology
from the fuel vendor to the licensee.
Within this report was
a
description of training administered to licensee
personnel
for the assumption
of reload analysis responsibility.
The inspectors
assessed
the adequacy of
the continuing training provided for responsible
personnel.
According to licensee
managements
fuel reload analyses will eventually
be
assigned
to one of the disciplines within the Nuclear
Fuel
Management
Department.
The four engineering disciplines comprising nuclear fuel
management
were fuel cycle services,
nuclear analysis.
safety analysis,
and
reactor engineering.
The licensee
had acquired
a reload analysis training
ackage
from the fuel vendor.
This training contained
21 separate
classroom
essons
that were administered
by fuel vendor instructors.
According to the
submitted report,
the total package
took 83 classroom
days to administer.
According to licensee
representatives,
personnel
were not administered
the
, total classroom
package.
but select
groups of lessons within the total
package.
The inspectors
asked if personnel
had initially been task
qualified'ccording
to the specific training received
and were informed that this was
not the case.
The inspectors
asked to see available documentation
indicating
the specific lessons that individuals had received.
They were informed these
records
had been lost during
a recent office relocation of most personnel
within the Nuclear Fuel
Management
Department.
The inspectors
asked
about
individual training record documentation
and were told that there were
some
training certificates within these records'ut
they were probably not
complete.
The licensee's
topical report stated that continuing training of personnel
with reload analysis responsibility
would be accomplished
by on-the-job
training.
The inspectors
attempted to determine the specific training
requirements.
On-the-job training requi rements for engineering
personnel
were
addressed
in "Engineering Personnel
Qualification
8 Training Program
Description." Revision 6.00.
This document described
the industry accredited
training program for engineering
support personnel.
A review of this document
indicated that reload analysis
personnel
assigned to the different groups
within nuclear fuel management
were subject to different training
requi rements.
Fuel cycle services
engineering,
nuclear analysis
engineering,
a'nd safety analysis
engineering
were described
as "non-accredited participant
groups."
Reactor engineering
was classified
as
an "accredited participant
group."
The inspectors
observed that the program description for accredited
or
non-accredited
groups did not include specific initial or continuing training
requirements
for fuel reload analysis activities.
This was pointed out to
licensee
representatives
who referred the inspectors
to previous Revision 5.00
of the training program description.
This revision contained
appendixes
for
nuclear analysis
and safety analysis
engineers
that specified requirements
and
frequency for reload analysis continuing training.
Reload analysis continuing
training was not addressed
for fuel services
engineers
and reactor engineers.
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The inspectors identified that the requirements
of Revision 5.00 were no
longer in effect,
and that Revision 6.00 to the training program description
had eliminated reload analysis continuing training requirements
for nuclear
and
sa fety ana lysi s engineers.
The inspectors
reviewed
a sample of personnel
training records
being
maintained
by supervisors of the safety analysis
and nuclear analysis
groups.
It was noted that one group's
records contained only database
printouts of
courses
completed.
while the other group's
records additionally contained
graded examinations
and course completion certificates.
The review of these
records did not indicate to the inspectors
what training constituted initial
reload analysis training and what constituted continuing training.
During the inspection, it was not apparent that reload analysis
team members
from specific groups within nuclear fuel management
requi red any task specific
training.
During interviews, the inspectors
became
aware that the safety
analysis
group had developed
personnel
qualification cards that included
performance
measures
for determining proficiency in reload analysis tasks.
However, these cards
had never
been
used
and it was not clear if they were to
be used in initial training, continuing training, or both.
Also, the
inspectors
could not determine if the cards
were intended to be used
by all
groups or just safety analysis
personnel.
According to a licensee-provided list of reload technology transfer
program
milestones,
the initial training package
procured
from the fuel vendor
had
been completed in April 1990.
Additionally. licensee
personnel
stated that
a
reload analysis continuing training program consisting of on-the-job training
and,
based
on
a three-year cycle.
had been
implemented
on completion of
initial training.
Based
on this information, the first 3-year cycle of
continuing training for personnel
who had completed initial training, should
have
ended in April 1993.
No documentation
could be provided in response to
the inspectors'equest
for information that could be used to assess
the
viability of the continuing training program for reload analysis.
Also, the
licensee did not have
a system for tracking or evaluating the progress of
continuing training related to reload analyses.
, The inspectors
could not fully evaluate the effectiveness
of the initial
training program
as records
were not available to indicate the specific
classroom training received
by specific analyses
team personnel.
From a
review of the program subject matter available
as classroom training'he
program appeared
to address all elements of reload analyses
adequately.
The
inspectors
determined that the present continuing training program for reload
analysis
was not
a viable program for the following reasons:
~
Overall program training requirements
had not been determined;
~
Training requirements
for team personnel within specific groups
had not
been determined:
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~
There was
no apparent
integration of reload analyses initial or
continuing training among the different groups that make
up
a reload
analysis
team;
and
~
Continuing training for reload analyses
personnel
was not routinely
tracked or evaluated.
The adequacy of continuing training for reload analyses
personnel
is an
Inspection Followup Item (528; 529; 530/9435-02).
)
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ATTACHMENT
1
Persons
Contacted
1. 1
Arizona Public Service
Com an
¹*J. Bailey. Vice President,
Nuclear Engineering
8 Projects
- R. Bandera,
Supervisor,
Nuclear Analysis,
Nuclear
Fuel
Management
¹ S.
Bauer.
Section Leader,
Licensing.
Nuclear Regulatory Affairs
¹*P. Crawley, Directors'uclear
Fuel
Management
- G. Duede
~ Senior
Engineers
Fuel Cycle Services,
Nuclear Fuel
Management
- D. Garchow, Director, Systems
Engineering
- B. Grabo,
Section
Leader,
Compliance,
Nuclear Regulatory Affairs
¹ J.
Gunn, Senior Engineer,
Reactor
Engineering,
Nuclear Fuel
Management
- C. Kar lson,
Senior
Engineer,
Nuclear Analysis, Nuclear Fuel
Management
¹~A. Krainik, Department
Leader,
Nuclear Regulatory Affairs
¹*D. Hedek, Senior Engineer,
Nuclear Assurance
Engineering
G. Michael'enior Engineer, Licensing'uclear
Regulatory Affairs
¹ H. Reid. Supervisor,
Safety Analysis, Nuclear Fuel
Management
K. Roberson.
Senior Engineer,
Compliance.
Nuclear Regulatory Affairs
¹*R. Rogalshi
~ Licensing Engineer,
Licensing,
Nuclear
Regulatory Affairs
¹*G. Shanker,
Department
Leader,
Nuclear Assurance
Engineering
- B. Thiele
~ Supervisor,
Reactor Engineering,
Nuclear Fuel
Management
S. Troisi
~ Manager.
Operations
Computer Systems
- N. Turley, Senior Engineer,
Licensing,
Nuclear Regulatory Affairs
¹*J.
Webb, Senior Engineer,
Safety Analysis, Nuclear Fuel
Management
1.2
NRC Personnel
- T. Gwynn, Director . Division of Reactor
Safety
- K. Johnston'enior
Resident
Inspector
- H. Wong. Chief. Reactor Projects
Branch
F
In addition to the personnel
listed above,
the inspector contacted
other
personnel
during this inspection period.
- Denotes those persons that attended the preliminary exit meeting
on
November 17.
1994.
¹Oenotes
those persons that attended
the exit meeting
on December
2,
1994.
2
EXIT MEETING
Interim exit meetings
were conducted
on November
17 and December
2, followed
by
a final telephonic exit meeting
on December
16,
1994.
During the meetings,
the inspectors
reviewed the scope
and findings of the report.
The licensee's
senior official expressed
a position on one of the apparent violations
presented
at the December
2,
1994.
meeting that consideration
should
be given
to its corrective actions
taken in response
to a prior violation involving
emergency diesel
generator operability.
In response.
the inspectors
requested
that the specifics of the licensee's
corrective actions
be telephoned
or sent
to the Regional office on the following Monday.
December
5,
1994.
The
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information was not sent.
but during the week of December
12,
1994.
a licensee
representative
contacted
the inspectors to inquire as to what information was
expected.
Subsequently,
on December
16.
1994, the licensee's
representatives
stated that this information would be sent to the Regional office for the
inspectors to review and that the information should reach the Regional office
on December
19 '994.
The information did not arrive,
and on
December
21,
1994 'he inspectors
informed the licensee's
representative
that
the inspectors'n
further reviewing this matter,
concluded that the apparent
violation that potentially related to the subject corrective action was
unwarranted.
During the inspections
the licensee's
representative
identified, in a general
manner, that various materials
reviewed by the inspectors
was proprietary.
The inspectors
stated that reasonable
judgement would be used to ensure that
no proprietary information was released
in the report.
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