ML17342A733
| ML17342A733 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 06/25/1987 |
| From: | Belisle G, Russell Gibbs, Moore L, Casey Smith, Wright R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17342A729 | List: |
| References | |
| 50-250-87-24, 50-251-87-24, NUDOCS 8707070227 | |
| Download: ML17342A733 (19) | |
See also: IR 05000250/1987024
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
'I01 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323
Report Nos.:
50-250/87-24
and 50-251/87-24
Licensee:
Florida Power and Light Company
9250 West Flagler Street
Miami,
FL
33102
Docket Nos.:
50-250
and 50-251
Facility Name:
Turkey Point
3 and
4
License Nos.:
and
Inspection
Conducted:
May 11-15,
1987
Inspectors: R:; Wrig
7
Date Signed
Date Signed
~
r8
~'.l5', ~ .
L.
R. Moorer/
Approved by:
G. A. Belisle, Chief
equal
ity Assurance
Programs
Section
Division of Reactor Safety
Date Signe'd
Date Signed
Date Signed
SUMMARY
Scope:
This special,
announced
inspection
was
conducted
in the
areas
of;
engineering
procedures
and controls for engineering
evaluations,
review and
assessment
of engineering
evaluations,
the training
and qualifications of the
performance
monitoring
section,
examination
of plant
operator
overtime
conditions,
and licensee
action
on previously identified inspection findings.
Results:
One violation was identified.
a707070227
870aS0
ADOCK 05000250
8
REPORT DETAILS
Persons
Contacted
Licensee
Employees
T.
S
AJ
- C
Ap
AW
- D
B.
wJ
- J
R.
- F
- J
- J
"R.
- p
- T
- C
F.
J.
- R.
- J
- W
A'J
- D
- p
"W.
A.
- G
W.
R.
- F
- C
S.
Abbatiello, Supervisor,
Quality Assurance
Performance
Monitoring
ection
Arias, Supervisor,
Regulatory
Compliance
Baker, Plant Manager
Banaszak,
Senior Electrical Engineer,
Juno
Power
Engineering
(JPE)
Bladow, Quality Assurance
(QA) Superintendent
Chancy,
Engineering
Manager
Crittendon, Site Engineering Organization
Dickey, Vice President,
Nuclear Operations
Donis, Site Engineering Supervisor
Farach,
QA Performance
Monitoring
Flugner,
Manager,
Technical
Licensing
Franklin, Engineering
Manager - JPE
Franzone,
Licensing Engineer
JPE
Gnecco,
Site Engineering Organization
Hart, Site Licensing Organization
Higgins, Technical
Licensing
Lead Technical
Engineer
Joseph,
Lead Civil/Structual Engineer - JPE
Kent, Manager,
Mechanical - Nuclear Engineering
Krenke,
QA Performance
Monitoring Section
Labarraque,
Technical
Department
Mende, Operations
Supervisor
Mendieta,
Manager,
Nuclear Services
Miller, Training Superintendent
Odom, Executive Assistant to Vice President - Nuclear Operations
Osborn, Site Engineer
Pace,
Nuclear Licensing Supervisor
Pike, Safety Engineering
Group
Ross,
QA Performance
Monitoring Section
Salamon,
Compliance
Engineer
Shook,
QA Performance
Monitoring Section
Stone,
QA Performance
Monitoring Section
Varona, Site Instrumentation
and Control
( IKC) Engineer
JPE
Wethy, Site Vice President
Wilker, Site Engineering Organization
NRC Resident
Inspectors
- D. Brewer, Senior Resident
Inspector
"K. Van Dyne, Resident
Inspector
"J. McDonald, Resident
Inspector
"Attended exit interview
LIST OF ABBREVIATIONS USED IN THIS
REPORT'P
CPWO
JCO
NRC
PC/M .
USQD
Administrative Procedure
Controlled Plant Work Order
Florida Power
and Light Company
Final Safety Analysis Report
Justification for Continued Operation
Nonconformance
Report
Nuclear
Regulatory
Commission
Plant Change/Modification
Quality Assurance
Quality Control
Systematic
Assessment
of Licensee
Performance
Social Security
Number
Turkey Point
Unreviewed Safety Question Determination
Exit Interview
The inspection
scope
and findings were
summarized
on
May 15,
1987, with
those
persons
indicated in paragraph
1 above.
The inspector described
the
areas
inspected
and
discussed
in detail
the
inspection
findings.
No
dissenting
comments
were received
from the licensees
Violation, Failure to follow the intent of Administrative Procedure
AP 0103.2
to
adequately
prevent
excessive
overtime
by
operator
personnel
in that
overtime
restriction
commitments
were
abused,
paragraph
8.
The licensee
did not identify as proprietary
any of the materials provided
to or reviewed by the inspector during this inspection.
Licensee Action on Previous
Enforcement Matters (92701)
(Cl osed)
Unresolved
Item
50-250,
251/87-09-01:
Excessive
Overtime
By
Operations
Personnel
This item was closed
during this inspection
because it was
upgraded
to
Violation 50-250,
251/87-24-01.
See
paragraph
8.
Unresolved
Items
Unresolved
items were not addressed
during this inspection.
Review of Engineering
Procedures
and Controls for Engineering
Evaluations
The controlling procedure
for the performance
of all design
and safety
analyses
by
Juno
Power
Engineering
i s
JPE-QI
3.2.
The
inspector
determined
that
depending
on
the
nature
of the activity involved,
the
following safety evaluations
are prepared
by JPE:
safety
evaluations
associated
with
plant
change
modification (PC/M) activities.
Justification
for'ontinued
Operations
(JCOs)
which
are
10'FR 50.59
nuclear
safety
reviews
of
discrepant
conditions
with
possible
associated
administrative
limitations
on
operations.
Controlled =Plant
Work Order
(CPWO)
design .control
documents
consisting
of
a
Part/Component
Compatability
Evaluation
(PCCE)
used for implementing
component substitution
changes.
The above safety evaluations
are performed to specifically address
changes
.to components
described
in the
FSAR, or to permi't plant operation with a
temporary deviation from FSAR commitments.
The inspector
determined
that
safety evaluations
are
also
performed for corrective actions
associated
with deficient
components,
and/or
discrepant
plant
conditions
not
described
in the
FSAR.
These
safety evaluations
are
performed
over
and
above
the
requirements
of
and
are
integral parts'f
the
corrective action program..
The inspector
reviewed the following procedures
to assess
the adequacy
of
the procedural
controls for performance of the above safety evaluations.
JPE-QI 3.2,
Design
and Safety Analyses
Performed
by JPE,
Revision 3.
I
Supplement
QI 3.2-1,
Functional
Design Reference
Guide,
Revision 3.
Supplement
QI 3.2-2,
Safety, Quality and Regulatory Reference
Guide,
dated 09-15-80.
Supplement
QI 3.2-3,
Design
Verification
Refer ence
Guide,
dated
09-29-83.
Review of the
above
documents
and
discussions
with licensee
personnel
revealed
that the administrative
controls delineated
in QI 3.2
and
its'ttachments,
are .applicable to all products
produced
by JPE, i.e.
PC/Ms,
JCOs
and safety evaluations.
The controls appear to be adequate
for PC/M
and
JCO processes.
Safety
Evaluations
associated
with significant
nonconformances
are
performed within the administrative
controls
of JPE-QI
3.2.
However,
because
specific guidance for
USQDs
had not been
established
for use
by
engineering
personnel
during the safety evaluation
process
performed to
disposition significant non-conformances,
requirements
for consideration
of the direct
effects
of
non-conforming
conditions
(hardware
and/or
software)
on safety-related
equipment function,
performance,
reliability,
and
response
time,
had
not
been
explicity defined.
Other
typical
considerations
normally included
during
the
safety
evaluation
process,
such
as the direct effect of non-conformance
single failure criteria,
separation
criteria',
seismic analysis, etc., defined
for thy
PC/M process,
had
not
been
established
in writing for evaluation of
.significant
nonconformances.
The
inspector
concluded
that while JPf-QI 3:2
can
be used to: control the performance
of the above
Safety. Evaluations, it is
gore
suitable
for Safety
Evaluations
prepared
in
support
of engineering
design activities.
The
inspector
reviewed
an
inter-office . correspondence
from'.
H.
Rogers,
Jr.
Power
Plant .EngineeAng,
$ubject:
Safety
Evaluations
for St.
Lucie
- and
Turkey .Point Plant,
File
GD-105,
dated
May 15,
1987.
This
document
addresses
the
evaluation
of the
conoseal
leak
on
Turkey Point Unit 4,
an'd
provides
dir~ctives
to
assure
improvement
in the
preparation,
review,
and
approval
of
safety
evaluations.'he
=directives
delineated .in
this
~ document ws'll be'incorporated
in procedure
QI 3.2 upon its next, revision.
The
requirements
to
be
-implemented
during
the
safety
evaluation
process
delineated
in the
above
inter"office correspondence
appear
to
be adequate.
However,
additional
improvement
in
the
preparation
of safety
evaluations
may
be
obtained
by
providing
prescriptive
guidance
to
engineering
personnel
for
USQDs.
Existing
administrative
controls
and
guidance
for
preparation
of safety
evaluations
are
not only delineated
in
QI 3.2,
but
are
also
addressed
in various
procedures
(eg.
QI
3. 1,
and
QI 3.3),
that
deal
with
the
engineering
design
program.
A
stand-alone
procedure,
intended specifically to
address
the
USQD 'process,-
would consolidate
these
guidelines,
and 'acilitate
preparation
of
adequate
nuclear
safety
evaluations
performed
pursuant
to
the
requirem'ents
of
or
10 CFR 50, Appendix B, Criterion 16.
The
inspector
determined
that
the
site
engineering
organization
is
presently
revising
their
procedures
that
delineate
the
administrative
controls
for
processing
NCRs.
The
inspector
conducted
interviews
with
engineering
personnel
and
reviewed
the following procedures
to
assess
the
nature of these
changes.
PTN-EP 2.3,
Processing
of Nonconformance
Report
(NCR), Revision
0
PTN-EP 2.7, Initial Engineering
Assessment
of Operability, Revision
0
Changes
being
made to the above procedures
are intended to more clearly define
the responsibilities, of the Site Engineering
Supervisor,
the Site
Engineering
Manager,
and the
Lead Discipline Engineer
concerning
the processing
of NCRs.
It is
the inspector's
understanding
that
these
changes
are
being
made
to
upgrade
the determination
process
of whether-or-not
safety
evaluations
are
required
for disposition of nonconformances.
Existing controls require 'that
this determination
be'ade
early during the initial engineering
assessment
of
the operability review process;
sometimes without relevant data to support the
decision
arrived at.
The inspector
was informed that the changes
are intended
to provide 'a
more controlled
decision
making
process
based
on
reviews of
pertinent
data
required
,to assess
a potential operability concern.
The
need
for
a
second
level
engineering
safety
assessment
to determine
the
safety
significance
of, the identified
nonconformance
is arrived at
based
on
the
previous evaluation .performed.
The
above
changes
are
an
improvement
in the disposition. of NCRs in that
effective follow-up corrective action is better
assured
for identified
deficiencies
having
operability
concerns.
These
changes
address
a
weakness
in the disposition of NCRs
from
a nuclear
safety
standpoint,
having
been
previously
documented
in
NRC
Inspection
Report
Nos.
50-250/87-02
and 50-251/87-02.
The
inspector
requested
information
concerning
the status
of procedure
PTN-EP
2.6,
which delineates
the controls for performing
engineering
safety assessments.
Engineering
safety assessments
represent
an activity
that follows the completion of the determination of potential operability
concern,
and
precedes
the
performance
of the
nuclear
safety
evaluation
addressed
in gI 3.2.
The inspector
was informed that licensee
management
has not yet decided
as to the
need to revise this procedure.
To assure
the performance
by JPE of safety
evaluations
for deficiencies
having operability
concerns,
a decision is required
by
FP5L management
regarding
the status
of procedure
PTN-EP 2.6.
This activity crosses
an
external
interface
between
the site
engineering
organization
and
JPE.
Therefore,
in order
to
implement
the
requi rements
of paragraph
3 of
W.
H. Rogers,
Jr.
correspondence
JPE-L-87-060,
administrative
controls
need
to
be established
to assure
performance
by JPE of required
safety
evaluations.
Within this area
no violations or deviations
were identified.
Review and Assessment
of Engineering
Evaluations
Selected
samples
of
PC/Ms,
CPWOs,
JCOs,
and
Safety
Evaluations
were
reviewed
by the inspectors
to assess
the adequacy of the safety evaluation
performed for each.
The samples
were selected
by each inspector according
to his technical
discipline;
the
nuclear
safety
significance
of the
activity involved;
and the
degree
of complexity of the activities to be
performed.
The following represents
the
sample that was reviewed during this effort:
PC/Ms Nos.
JCO Nos.86-005
86-071
86-014
86-136
86-145
86-053
86-102
86-052
86-060
87-098
87"093
87-008
87-037
87-113
J PE- L-86-113
J P E-L-86-108
JPE-L-86-050
JPE-L"86-092
JPE-L-86-049
JPE-M-86-017
JPE-M-86-077
JPE-M"86-060
JPE-M-86-091
Safet
Evaluations
Nos.
CPIIIO Nos.
JPE-L-86-50
JPE"L-86-69
JPE-L-86-93
JPE-L-86-111
JPE-L-86"49
JPES-E-86-814
J P ES-E-86-1008
PTPO-86-1474-E-1
86-245
86-035
86-086
86-125
86-062
86-054
87-071
The
documented
basis
for the
USED was
assessed
by the inspectors
with
regard
to the effect of the
change
(or activity)
on
a safety-related
equipment
function,
performance,
reliability,
and
response
time if
applicable.
Typical
engineering
design
requirements
of single failure
criteria,
separation
criteria,
and
seismic
requirements
were
also
considered
during
the
evaluation
process.
The
inspectors
conducted
detailed discussions
with the cognizant
engineers
to clarify issues
that
were raised during the evaluation,
and to obtain additional
documentation
. equi red to resolve questions
and/or concerns.
The
inspectors
did not identify any deficiencies
with the engineering
evaluation
process
performed for the selected
sample
reviewed.
The
inspectors
performed
a
review
and
assessment
of the
adequacy
with
which site engineering
organization
personnel
disposition Operations
NCRs.
The data
base
used
during this assessment
were the
NCR Logs for 1986 and
1987,
and
consisted
of
a total
of
420
NCRs.
Because
of
the brief
descriptions
of the
non-conformances
contained
in these
logs,
select
samples
could not
be easily obtained
to assure
review and
assessment
of
substantive
issues.
A statistical
analysis
of the
data
was therefore
conducted.
A procedure
employing discovery
sampling
techniques
was
used
to determine
the
sample
size to
be selected
on
a
random basis,
and to
assess
an error rate within the overall population.
Reference
material
used
during this effort was
the
"Handbook of Sampling for Auditing and
Accounting-2nd Edition," by Herbert Hawkins, Published
by McGraw Hill Book
Company.
The objective of discovery
sampling is to provide
a specified
assurance
of
seeing at least
one example of an attribute if its rate of occurrence
in
the population is at or greater
than
a specified rate.
This procedure
is
appropriate
when
the
occurrence
rate
is quite
low, yet
a
specified
probability value is, desired to assure
that the
sample is large
enough to
expect the identification of at least
one example.
A description of the attributes
as defined
by precise characteristics
has
.to
be
made.
The inspectors
selected
.the following as attributes
to be
assessed
during this effort:
Adequacy of i'dentification of the root cause
of the
non-conforming
condition.
adequacy of developed corrective action plan.
Adequacy of post-installation/post-modification. test requirements
and
test acceptance
criteria.
- Adequacy of test results if available.
The
inspectors
assumed
an error rate of .5% in engineering
evaluations
performed
by site engineering
personnel
for dispositioning
NCRs.
An error
'was defined
as identifed deficiencies
in any of the above attributes.
A
confidence
level (probability value) of
99% was selected
to assure
the
identification of at least
one
example, if indeed
the error rate
was as
high as
assumed.
Based
on the reference
text used,
for a population of
420,
a confidence
level of 99%,
and
an error rate of 5%, the
sample
size
was determined to be 84 from Appendix J,
page
472.
A
random
selection
of 84
was
obtained
from the
population
and
supporting
documentation
was provided by licensee
management
in accordance
with the inspectors'equests.
NCRs having little or
no involvement
by
engineering
personnel
for disposition were replaced -by others
selected
at
random.
The inspectors
reviewed
the documentation
to assess
the
adequacy
of the
root cause
determination.
NCRs perceived
by the inspectors
as requiring
in-depth
review
because
of
a
concern,
were
given
special
attention.
Developed
corrective
action
plans
were
reviewed
and discussed
with the
cognizant
engineers
to clarify issues
of nuclear
safety
significance,
post-installation/post
modification test
requirements
and test results.
Additional documentation
required
by the inspectors
to resolve
issues
and
concerns
was also provided by the cognizant engineers..
Based
on the review and assessment
of the 84 NCRs, the inspectors
did not
identify an example that indicated inadequacies
in any of the attributes
used
as parameters
for assessment.
This result
may be interpreted to mean
that there is
a
99% probability that the error rate in engineering
reviews
performed
to disposition
is
less
than
5%.
Additionally,
from
Appendix J,
page
470, it can also
be
seen that there is
a
90% probability
that the error rate is less than
2.5%.
Within this area
no violations or deviations
were identified.
Training and Qualifications of the Performance
Monitoring Section
(PMS)
A
recent
NRC
Inspection.,
documented
in
NRC
Inspection
Report
Nos.
50-250,
251/87-09,
raised
a
concern
over
the
qual'ification of
personnel
in
the
newly established
PMS of the
Turkey Point Quality
Assurance
Organization.
This concern specifically addressed
an
issue of
personnel
performing inspection/survei llances in areas
where they were not
qualified.
Lack of training, for
PMS personnel
'in the. new standardized
Technical
Specifications
currently befng
used
on
a trial basis
(pending
NRC. approval),
was
also
addressed.
A portion of this
inspection
was
specifically
conducted
to follow-up on
the
concerns
raised
during
NRC
Inspection
50-250,
251/87-09.
.-The 'Quality
Organization
at
Turkey
Point
consists
of:
a
(construction)
QC group which is involved primarily in plant modifications
which reports to the Project Site Manager,
an operations
QC group which is
primarily involved in operations
type surveillance
which reports to the
Plant Manager,
and the Site Quality Assurance
Organization
which reports
to the
Site
QA Superintendent
and ultimately through
the Corporate
Director to the Group Nuclear Vice President.
The Site
QA organization is
divided into two sections:
The Regulatory
Compliance Section
and the
PMS.
The Regulatory
Compliance Section is responsible
for the site auditing in
order
to
meet
Appendix
B requirements.'he
PMS was created
during
1986 to establish
a
program for real-time in-plant monitoring of
performance,
specifically aimed at improving .the Plant's
SALP ratings
by
monitoring weak areas.
The stated
purpose
of this section
is to provide
for objective
evaluation
and quantitative verification that
designated
activities are being performed
in accordance
with specific technical
and
quality related
requirements.
In addition,
the
program also
provides
a
method for reporting pertinent
observations
of
a
more
subjective
or
qualitative nature,
which may be identified by
QA personnel
as they carry
out their assigned
functions.
The
section's
activities
are
divided
between
the functional
areas
of
maintenance,
operations
and surveillance/testing.
The section consists of
a supervisor
and nine people divided equally to cover the
above mentioned
functional areas.
Inspection of the activities of the
PMS consisted of interviews with seven
of the ten individuals in the group;
a review of the current
educational
backgrou'nds,
previous
certifications
held
and their
work experience.
Training provided to these
personnel
by the licensee
was also
reviewed.
This information was compared to recent work assignments
and. the following
facts were determined
by the inspection:
All of the
personnel
currently in the
section,
except
one,
have
Bachelor of Science
degrees
in scientific or engineering fields from
a college or university.
One individual. possesses
a Masters degree.
Three of the individuals have professional
engineer certificates.
. Three
of the
individuals
have
had
a
Reactor
Operator
or Senior
'eactor
Operator
license
issued
by the
NRC.
Two other individuals
have
had
a significant amount of plant
systems
training with one
being qualified as
a Shift Technical
Advisor.
One other individual
is currently attending reactor operator training.
The total nuclear industry experience for the group is 112 years for
an 'average
of, 11.2 years
per individual'with the least
amount of
experience
being five years.
Comparison
of
recent
work
assignments
with
the
individual's
background,
revealed
that
work assignments
correspond
with the
individual's background.
Re'view of the status of completion of licensee
provided training for
the
individuals
in
the
section,
revealed
that
six of
the Aen-
individuals
have
completed
the training
program.
The other
four
individuals have completed at least
75 percent of the training.
Some
'eakness
in the technical
content of this training program was noted,
but,. the licensee
has recognized this problem and is taking action to
improve the training.
This i s reflected
in the
1987 training plan
(dated
February 6,
1987)
which includes training for personnel
in
plant'ystems,
technical
specifications,
chemistry,
security
and
'health physics..
In conclusion,
although
some
weakness
was
noted
in licensee
provided
training,
the
individuals in the
appear
to
be well qualified to
perform the activities assigned
to the group.
8'.
Examination of Plant Operator Overtime Conditions
The information tabulated
on the last page of this report is
a chronology
of NRC policy statements
issued
concerning
nuclear
power plant (operator)
shift working hours
and the licensee's
procedural
and proposed technical
specification
commitments
.to implement this 'policy.
As stated
in paragraph
6d. of .NRC Inspection
Report Nos.
50-250,
251/87-09
this item was
made
an unresolved
item for the following reasons:
(1) after cursory review it was not clear to the inspector whether or not
the licensee
was in full compliance in regards to NUREG-0737 requirements
concerning overtime, (2) nor that the administration of overtime.is in full
accordance
with Administrative Procedure
AP-0103.2.'n
regards
to question
( 1) above,
examination of the current revision of
AP-0103.2
(9/17/86)
which
has
been
in effect
since
February 9,
1981,
indicates
agreement
with NUREG-0737 operator overtime requirements
except
that authorization to deviate
from overtime requirements
was granted to a
level below the plant manager's
deputy (the Operations
Supervisor) at TP.
Discussions
with the Operations
Supervisor
and examination
of pertinent
related
'documentation
revealed
he
has
authorized
deviations
from the
- overtime
requirements.
FPL's'P
proposed
Technical
Specification
(TS)
10
change
(submitted 9/30/86)'oes
not permit deviation authorization
below
the
plant
manager" s level.
If the
licensee
desires
to
extend
this
authority
to its
Operations
Superintendent,
it
should
amend
this
'TS.change
to say so.
-The inspector
compared
the
1986 bi-weekly time sheets
for five reactor
operators
(Operator
A, B,
C,
D, and
E) against
the overtime restrictions
. of AP-0103.2 to resolve
question
(2) above.
Two of the five operators
. were 'found to have 'violated criterion (1)
and (6)
on the chart in that
,they,'orked
14 straight
hours= and
had
bi eaks, of less
than
12
hours
between
work periods for the following time frames:
Operator
A
Operator.
C
Worked
14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> straight 5/1-2/86
10'hour .break 5/1/86
Worked
14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> .straight 7/22/86
'9~ hour, break 7/22-23/86
- .It
should
be
noted
that
although
the
licensee's
more
conservative
self-imposed
requirements
(AP-0103.2)
were violated
in this
case,
the
above instances
would not be violations in accordance
with current Generic
Letter
(GL)
82-12-02
requirements
or
th'e
proposed
TS
change
requirements.
Four of the five operators.
examined
were found to have worked more than
72
hours (criterion (3)) in the following 7-day period:
Operator
A
Operator
B
Operator
D
Operator,
E
4/28
5/4/86
7/21"27/86
8/7-13/86
8/17-23/86
7/23"29/86
7/20-26/86
4/5-11/86
4/13-19/86
74 hrs
86', hrs
85~ hrs
88 hrs
84 hrs
78 hrs
93 hrs
85.5 hrs
Three of the five operators
examined
were found to have worked more tPan
14 consecutive
days
without having
two consecutive
days off (criterion
(4)) during the following periods:
Operator
A
Worked
16 consecutive
days
between
7/21
8/5/86,
had
8/6/86 off, then worked 7 more days before getting
2-consecutive
days off
Operator
D
8/2-26/86
25 days
Operator
E
3/17
4/30/86
45 days
Similar restrictions
do not exist under
NRC GL 82-12 nor FPL's proposed
TS
change;
however,
the
above
examples
violate
the
licensee's
current
approved
commitments in AP-0103.2.
The
inspector
pointed
out
instances
where
some
of
FPL's
current-
self-imposed
commitments
(AP-0103.2)
have
been violated but, in fact, are
more conservative
than current
NRC GL 82-12 requirements.
However, if FPL
were operating
Under their proposed.
TS change
(which meets
requirements
of
GL,82-12)
commitments
during the
same
1986 timeframe,
there
would have
been
at
leas't
15 additional
examples
of violations occurring
by these
same
5 operator s for working
more
than
24-hours
in
a
48
hour period
, (criterion (2)).
'FPL has
committed to the
NRC opera'tor
overtime policy which states
that
operator
overtime shall
not be routigely scheduled
to compensate
for an
inade'quate
number of J icensee
personnel
to meet shift crew requirements.
The policy -also states
that in t;he event
overtime
must
be worked, due to
'.unanticipated
or
unavoidable
circumstances
. this
overtime
shall
be
controlled
by specified-overtime
restrictions.
Based
on
the
amount
of
actual
overtime
hours
worked during
1986,
and
the
number
of overtime
restric'tion violations identified for the five randomly
chosen
operators
examined, it .appears .that
management
may not fully understand
the
, intent of the
NRC policy to control operator
overtime
and
has
not been
actively involved in preventing the abuse of overtime restrictions.
This
formerly identified
unresolved
'item
50-250,
251/87-09-01,
Excessive
Overtime
By Operations
Personnel
is closed
but is upgraded
to violation
250,
251/87-24-01,
Failure
to
Follow Procedures
to Adequately
Control
Excessive
Overtime
By Operators
Personnel.
Licensee Action on Previously Identified Inspecti'on
Finding
(Closed) Inspector
Follow-up Item (250, 251/87-02-01)
Completion of Design
Control Procedures
JPE-AP-3. 1'1 and ASP-4.
FP&L's letter L-86-389 described corrective actions to be taken for design
control
10 CFR 50.59 evaluatioris,
and timeliness of corrective actions.
A
review of selected
elements
of the
developed
corrective
action
plans
revealed
that the
above
procedures,
which delineate
requirements
of the
design
program,
had not been completed.
Subsequent
inspection of licensee
actions
in this areas
has identified progress.
Procedure
ASP-4,
Change
Request
Notice Control,
was approved
on February
24,
1987.
Additionally,
from discussions
with the -licensee staff; apparent
progress is being
made
towards
the
issue
and
approval .of
JPE-AP-3. 11.
Based 'n
review of
objective
evidence
and
discussions
with licensee
staff, this
issue is
closed.
(Closed)
Inspector
Fol low-up
Item
(250,
251/85-18-01)
Programmatic
Deficiencies within PC/M Program.
This
item
addressed
weaknesses
in programmatic
controls
from initial
identification of a station
problem to development
and implementation of a
PC/H package.
Subsequent
corrective actions
by licensee
management
have
resulted
in program
documents
that
have defined
the responsibilities
of
various
organizations
involved
in
the
PC/N
program.
Additionally,
external
interfaces
between
organizations
performing work affecting
the
12
qual,ity of design,
have
been identified in writing.
The capabi-lity to
relate
a final design
back to its design input has
also
been
established
within the
program
documents.
This item is closed
based
upon review of
the following approved
procedures.
AP 0190. 15,
Plant
Changes
and Modifications (PC/M), dated April 7,
1987.
AP 0190.88,
Request for Engineering
Cost Estimate
(REE)-Preparation,
Review,
and Approval, dated
November 25,
1986.
AP 0190.87,
Project
Review Board, dated
November 25,
1986.
AP 0190.84,
Request
for Engineering
Assistance
(REA)-Preparation,
Review,
and Approval, dated
November 25,
1986.
AP 0190.81,
Change
Review Team (CRT), dated
November 25,
1986.
0190.82,
Request
for Technical
Assistance
(RTA)-Preparation,
Review,
and Approval, dated April 7,
1987.
PC/M Review Guidelines,
Form I.
PC/M Technical
Review Guidelines,
Form II.
NRC POLICY ON
PLANT OPERATOR
ERTIME RESTRICTIONS
VRS
TURKEY POINT
PLANT COMMITMENTS
NRC Interim
Criteria For
Shift Staffing
7/31/80
FPL Administrative
Procedure
AP 0103.2
11/13/80
NRC THI
Action Plan
Requirements,
NUREG-0737 11/80
FPL Administrative
Procedures
AP 0103.2
(2/9/81 - 9/17/86)
6/15/82
FPL Proposed
Technica
I
Specification
Change 9/30/86
(1)
12 hrs Hax.
straight time
Same
as (1)
(2) 24 hrs Max./48 hr
Same
as (2)
period
Same
as (1)
Same
as (1)
(6) Break of at
Same
as (6)
at least
12 hrs
between work
periods
(8)
16 hrs Max.
straight time
(9)
Same
as (2),
plus break of
at least
8
hours
between
work periods
Same
as (8)
Same
as (9)
( 3 )
72 hrs Max. /7-day
period
(4)
14 Consecutive
days/ 2-days off
Same
as (3)
Same
as (4)
Same
as (3)
Same
as (4)
Same
as (3)
Same
as (4)
Same
as (3)
No Criteria Stated
Same
as (3)
No Criteria
Stated
(5)
Deviation
authorization
plant manager or
higher levels of
management
Deviation authorization
not addressed
(7) Deviation
Ops.
Supervisor,
Ops.
authorization
Superintendent
or
plant manager,
Plant Manager
his deputy, or
higher levels
of management
Same
as (7)
Same
as (5)
(10) Except during
Saine
as (10)
extended
4
shutdown
periods,
overtime
should
be
considered
on
individua I
basis
and not
for entire
staff on
a
shift
Additiona I Notes:
Letter dated
October 20,
1981,
form NRR, Division of'icensing to FPLs Vice President
states
that after review FPLs
policy on overtime restriction for Turkey Point was found in agreement with NUREG-0737 and their policy was
deemed
acceptable.
Letter (L-82-436) dated October 12,
1982,
from FPL Vice President
to
NRR, Division of Licensing states
that it is
position that the overtime requirements
are adequately
enforced
by incorporated Administrative Procedures
and
amending
the Technical Specifications
is unnecessary
at this time.